WorldWideScience

Sample records for enhancing npp safety

  1. Safety enhancement in NPP Bohunice

    International Nuclear Information System (INIS)

    Lipar, M.; Mihalik, M.

    1997-01-01

    The upgrading and safety enhancement of both the Bohunice V-1 and V-2 reactors is described in detail. The total estimated cost of the gradual reconstruction of these two units during 1996 to 1999 is 180 mil. US dollars. For the 1995 to 1997 period, the actions common for both units include a quality assurance programme, a personnel training programme, installation of a multifunction simulator, implementation of symptom-oriented operation procedures, installation of diagnostic systems, of a site security system, and of a teledosimetric system. At present, the main maintenance tasks are: to carry out major repair of units, to remedy service interruptions, to enhance equipment service availability, to enhance the technical level of corrective actions at equipment. Investment into maintenance level upgrade has grown from 7.5 mil. Slovak crowns in 1994 to estimated 32 mil. in 2000. The partners of international cooperation are mentioned. (M.D.)

  2. The enhancement of Ignalina NPP in design and operational safety

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1999-01-01

    Enhancement of Ignalina NPP design include: core design improvements; fuel channel integrity (multiple pressure tube rupture); improvements of shutdown systems; improvements of instrumentation and control devices; containment strength and tightness; design basis accident analysis; improvements of safety and support systems; seismic safety enhancement; Year 2000 project; cracks in pipes. Enhancement of operational safety includes: quality assurance; configuration management; safety management and safety culture; emergency operating procedures; training and full scope simulator; in-service inspection; fire protection and ageing monitoring and management

  3. NPP Mochovce nuclear safety enhancement program

    International Nuclear Information System (INIS)

    Cech, J.; Baumester, P.

    1997-01-01

    Nuclear power plant Mochovce is currently under construction and an extensive nuclear safety enhancement programme is under way. The upgrading and modifications are based on IAEA documents and on those of the Nuclear Regulatory Authority of the Slovak Republic. Based on a contract concluded with Riskaudit from the CEC, safety examinations of the Mochovce design were performed. An extensive list of technical specifications of safety measures is given. (M.D.)

  4. Enhancing NPP Safety Through an Effective Dependability Management

    Energy Technology Data Exchange (ETDEWEB)

    Vieru, G., E-mail: g_vieru@yahoo.com [AREN, Bucharest (Romania)

    2014-10-15

    Taking into account the importance of the continuous improvement of the performance and reliability of a NPP and practical measures to strengthen nuclear safety and security, it is to be noted that a good management for a nuclear power reactor involves a ''good dependability management'' of the activities, such as: Reliability, Availability, Maintainability (RAM) and maintenance support. In order to evaluate certain safety assessment criteria intended to be applied at the level of the nuclear reactor unit management, equipment dependability indicators and their impact over the availability and reactor safety have to be evaluated. Reactor equipment dependability indicators provide a quantitative indication of equipment RAM performances (Reliability, Availability and Maintenance). One of the important benefits of maintenance and failure data gathering is that it can be used as a support of probabilistic safety assessment (PSA). Also, a good dependability management implementation may be used to complement reactor level unit performance indicators in the field of safe operation, maintenance and improving operating parameters, as well as for Strengthening Safety and Improving Reliability of a NPP. This paper underlines the importance of nuclear safety and security as prerequisites for nuclear power. In addition, it demonstrates how different technical aspects, through implementation of a good dependability management, contribute to a strengthened safety and an improvement of availability of the NPP through dependability indicators determination and evaluation. (author)

  5. Training system enhancement for nuclear safety at PAKS NPP

    International Nuclear Information System (INIS)

    KIss, I.

    2000-01-01

    Paks Nuclear Power Plant is the only commercial nuclear facility in Hungary, which has been operational since 1982. The over 15 years operation of the plant can from all aspects be considered as a success, to which the well qualified, competent staff significantly contributes. Like other N-plants, Paks NPP is also exposed to major challenges due to plant ageing and changes in circumstances that affect the operation. The management focusing on maintaining nuclear safety launched an overall programme to upgrade quality of personnel training and to improve its infrastructure. Though this programme has successfully finished with visible proofs, further actions to develop a reconsidered human resource policy is needed so that the plant would successfully stand against the challenges of the 21. century. (author)

  6. Mochovce NPP safety measures evaluation from point of view of operational safety enhancement

    International Nuclear Information System (INIS)

    Cillik, I.; Vrtik, L.

    2000-01-01

    Mochovce NPP consists of four reactor units of WWER 440/V213 type and it is located in the south-middle part of Slovakia. At present first unit operated and the second one under the construction finishing. As these units represent second generation of WWER reactor design, the additional safety measures (SM) were implemented to enhance operational and nuclear safety according to the recommendations of performed international audits and operational experience based on exploitation of other similar units (as Dukovany and J. Bohunice NPPs). These requirements result into a number of SMs grouped according to their purpose to reach recent international requirements on nuclear and operational safety. The paper presents the bases used for safety measures establishing including their grouping into the comprehensive tasks covering different areas of safety goals as well as structural organization of a project management of including participating companies and work performance. More, results are given regarding contribution of selected SMs to the total core damage frequency decreasing. (author)

  7. New generation main control room of enhanced safety NPP with MKER reactor

    International Nuclear Information System (INIS)

    Golovanev, V.E.; Gorelov, A.I.; Proshin, V.A.

    1994-01-01

    Russia is planning to begin the gradual substitution RMBK NPP units, whose resources were worked out itself, to NPP units with a 800 MW multiloop boiling water power reactor (MKER-800) enhanced safety at next ten-year period. Main drawbacks of RBMK Reactor were completely removed in design of MKER-800 reactor. Moreover some special decisions were made to give MKER-800 self-safety properties. The proposed design of the MKER-800 enhanced safety reactor is not only fully free from the drawbacks of the RBMK reactors, but also show a number of advantages of channel-type reactors. This Paper presents some preliminary proposals of MCR Design, that developed Research and Development Institute of Power Energy (RDIPE). 6 refs, 2 figs

  8. Safety enhancement concept for NPP of new generation with VVER reactors

    International Nuclear Information System (INIS)

    Bezlepkin, V.; Kukhtevich, I.; Semashko, S.; Svetlov, S.; Solodovnikov, A.

    2004-01-01

    With the present day conditions, in order to successfully promote new NPP designs in the electric power markets, it is necessary to ensure enhanced technical/economic performances provided that international safety requirements are properly adhered to. When compared with high-powered nuclear power plants, NPP VVER-640 design (medium powered) possesses a number of advantages for the regions with undeveloped energy systems. Reduced specific energy intensity of the core adopted in this type of reactor allows to ensure the emergency cooldown of the reactor plant by passive means and to minimize the 'human factor' risk and external effects and provide sound substantiations as to how to retain corium inside RPV in case of severe accidents. At the same time, high-powered NPPs seem to be promising for regions with developed energy systems. Among such designs, NPP VVER-1000 and VVER-1500 designs are the most desirable. Configuration of new generation NPP with VVER-1500 is to be selected based on the gained experience in designing NPPs of previous generations considering the latest safety requirements and situation in the domestic and global energy markets for the time being and in the short run. Recent IAEA publications and latest EUR requirements insist that the following key safety indices should be established for new NPP designs: - aggregated frequency of core melting is 10 -6 (1/year); - frequency of maximum accident release is 10 -7 (1/year). To meet the aforementioned criteria, it is necessary to implement some safety assurance principles recommended by IAEA (in-depth defence, single failure, redundancy, diversity, etc.), application of deterministic and probabilistic methods for selection of safety assurance activities and means and use of reasonable combination of active and passive systems. Application of VVER-640 concept to high-powered NPPs seems to be a formidable task due to a number of reasons, namely, it is quite difficult to carry out cooldown process

  9. Safety culture at Mochovce NPP

    International Nuclear Information System (INIS)

    Markus, Jozef; Feik, Karol

    2002-01-01

    This article presents the approach of Mochovce NPP to the Safety culture. It presents activities, which have been taken by Mochovce NPP up to date in the area of Safety culture enhancement with the aim of getting the term into the subconscious of each employee, and thus minimising the human factor impact on occurrence of operational events in all safety areas. The article furthermore presents the most essential information on how the elements characterising a continuous progress in reaching the planned Safety culture goals of the company management have been implemented at Mochovce NPP, as well as the management's efforts to get among the best nuclear power plant operators in this area and to be an example for the others. (author)

  10. Operation safety at Ignalina NPP

    International Nuclear Information System (INIS)

    Zheltobriukh, G.

    1999-01-01

    An improvement of operational safety at Ignalina NPP covers: improvement of management structure and safety culture; symptom-based emergency operating procedures; staff training and full scope simulator; program of components ageing; metal inspection; improvement of fire safety. The first plan of Ignalina NPP Safety culture development for 1997 purposed to the SAR recommendation implementation was prepared and approved by the General Director

  11. AGNES - safety reassessment of Paks NPP

    International Nuclear Information System (INIS)

    Gado, J.

    1995-01-01

    The main goal of the AGNES (Advanced General and New Evaluation of Safety) project for the reassessment of the safety of Paks Nuclear Power Plant, Hungary, was to improve the safety culture of the technology at Paks. A report was prepared on the reassessment of the Paks NPP safety. The analysis was divided into four groups: systems analysis, analysis of design basis accidents, severe accident analysis, and level 1 probabilistic safety analysis. Proposed safety enhancement measures are discussed. (N.T.)

  12. Regulatory aspects of NPP safety

    International Nuclear Information System (INIS)

    Kastchiev, G.

    1999-01-01

    Extensive review of the NPP Safety is presented including tasks of Ministry of Health, Ministry of Internal Affairs, Ministry of Environment and Waters, Ministry of Defense in the field of national system for monitoring the nuclear power. In the frame of national nuclear safety legislation Bulgaria is in the process of approximation of the national legislation to that of EC. Detailed analysis of the status of regulatory body, its functions, organisation structure, responsibilities and future tasks is included. Basis for establishing the system of regulatory inspections and safety enforcement as well as intensification of inspections is described. Assessment of safety modifications is concerned with complex program for reconstruction of Units 1-4 of Kozloduy NPP, as well as for modernisation of Units 5 and 6. Qualification and licensing of the NPP personnel, Year 2000 problem, priorities and the need of international assistance are mentioned

  13. Safety aspects of NPP ageing

    International Nuclear Information System (INIS)

    1995-01-01

    Preparation of safety practices on assessment and management of aging of major NPP components important to safety, CRP on management of aging of concrete containment buildings, CRP on management of aging of in-containment instrumentation and control cables are outlined

  14. Enhancing operational safety

    Energy Technology Data Exchange (ETDEWEB)

    Wiebe, J S

    1997-09-01

    The presentation briefly considers the following aspects concerning enhancing operational safety of NPP: licensed control room supervision, reactivity changes, personnel access to control room, simulator training.

  15. The possibilities of applying a risk-oriented approach to the NPP reliability and safety enhancement problem

    Science.gov (United States)

    Komarov, Yu. A.

    2014-10-01

    An analysis and some generalizations of approaches to risk assessments are presented. Interconnection between different interpretations of the "risk" notion is shown, and the possibility of applying the fuzzy set theory to risk assessments is demonstrated. A generalized formulation of the risk assessment notion is proposed in applying risk-oriented approaches to the problem of enhancing reliability and safety in nuclear power engineering. The solution of problems using the developed risk-oriented approaches aimed at achieving more reliable and safe operation of NPPs is described. The results of studies aimed at determining the need (advisability) to modernize/replace NPP elements and systems are presented together with the results obtained from elaborating the methodical principles of introducing the repair concept based on the equipment technical state. The possibility of reducing the scope of tests and altering the NPP systems maintenance strategy is substantiated using the risk-oriented approach. A probabilistic model for estimating the validity of boric acid concentration measurements is developed.

  16. Operational experience review and methods to enhance safety and reliability in the NPP-Leibstadt (KKL)

    Energy Technology Data Exchange (ETDEWEB)

    Haeusermann, R [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland)

    1997-10-01

    In the nuclear community it became clear that an integrated feedback system of operating experience must also include the unsuccessful results. The deviations, expected to achieved performance are analysed to the failure mode and its effect. KKL has lowered the number of safety significant events since commercial operation started. The thoroughness of the review/analysis of the events has increased with high priority set to human factor induced events in operation and maintenance. Since the participation of the author in the ASSET-Mission in Smolensk in 1993, KKL introduced the ASSET-Root-Cause method and has supplemented it by the HPES (Human Performance Enhancement System). 4 refs, 6 figs.

  17. Operational experience review and methods to enhance safety and reliability in the NPP-Leibstadt (KKL)

    International Nuclear Information System (INIS)

    Haeusermann, R.

    1997-01-01

    In the nuclear community it became clear that an integrated feedback system of operating experience must also include the unsuccessful results. The deviations, expected to achieved performance are analysed to the failure mode and its effect. KKL has lowered the number of safety significant events since commercial operation started. The thoroughness of the review/analysis of the events has increased with high priority set to human factor induced events in operation and maintenance. Since the participation of the author in the ASSET-Mission in Smolensk in 1993, KKL introduced the ASSET-Root-Cause method and has supplemented it by the HPES (Human Performance Enhancement System). 4 refs, 6 figs

  18. Regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning

    International Nuclear Information System (INIS)

    Lipar, M.

    1997-01-01

    A review of regulatory requirements for demonstration of the achieved safety level at the Mochovce NPP before commissioning is given. It contains licensing steps in Slovakia during commissioning; Status and methodology of Mochovce safety analysis report; Mochovce NPP safety enhancement program; Regulatory body policy towards Mochovce NPP safety enhancement; Recent development in Mochovce pre-operational safety enhancement program review and assessment process; Licensing steps in Slovakia during commissioning

  19. Safety culture in Ignalina NPP, regulatory view

    Energy Technology Data Exchange (ETDEWEB)

    Maksimovas, G [VATESI (Lithuania)

    1997-09-01

    The presentation describes how success on the way to a high level Safety Culture in Ignalina NPP may be achieved by daily, well motivated activities with good attitude and proper management participation, ensuring the development and proper implementation of Safety Culture principles within the activities of Operational organization of Ignalina NPP.

  20. Safety culture in Ignalina NPP, regulatory view

    International Nuclear Information System (INIS)

    Maksimovas, G.

    1997-01-01

    The presentation describes how success on the way to a high level Safety Culture in Ignalina NPP may be achieved by daily, well motivated activities with good attitude and proper management participation, ensuring the development and proper implementation of Safety Culture principles within the activities of Operational organization of Ignalina NPP

  1. Indicators to monitor NPP operational safety performance

    International Nuclear Information System (INIS)

    Gomez-Cobo, Ana

    2002-01-01

    Since December 1995 the IAEA activities on safety performance indicators focused on the elaboration of a framework for the establishment of an operational safety performance indicator programme. The development of this framework began with the consideration of the concept of NPP operational safety performance and the identification of operational safety attributes. For each operational safety attribute, overall indicators, envisioned as providing an overall evaluation of relevant aspects of safety performance, were established. Associated with each overall indicator is a level of strategic indicators intended to provide a bridge from overall to specific indicators. Finally each strategic indicator was supported by a set of specific indicators, which represent quantifiable measures of performance. The programme development was enhanced by pilot plant studies, conducted over a 15 month period from January 1998 to March 1999. The result of all this work is compiled in the IAEA-TECDOC-1141, to be published shortly. This paper presents a summary of this IAEA TECDOC. It describes the operational safety performance indicator framework proposed and discusses the results of and lessons learned from the pilot studies. Despite the efforts described, it is clear that additional research is still necessary in areas such as plant-specific adaptation of proposed frameworks in order to suit individual data collection systems and plant characteristics, indicator selection, indicator definition, goal setting, action thresholds, analysis of trends, indicator display systems, analysis of overall safety performance (i.e., aggregation or combination of indicators), safety culture indicators, qualitative indicators, and use of additional indicators to address issues such as industrial safety attitude and performance, staff welfare, and environmental compliance. This is the rationale for a new IAEA Coordinated Research Project on 'Development and application of indicators to monitor NPP

  2. Completion of the VVER 440/213 NPP Mochovce incorporation enhanced safety features

    International Nuclear Information System (INIS)

    Charbonneau, S.; Eckert, G.

    1996-01-01

    The cooperation between the western countries and the countries of ex-eastern block in the field of nuclear safety is recent and still limited. The main reasons for this situation are limited or non existent capabilities of these countries for financing as well as non acceptable legal conditions concerning the third party nuclear liability in this part of Europe. Nevertheless, Framatome and Siemens associated in the consortium named EUCOM, have signed in April 1996 the contract of about 100 million US dollars with Slovak electricity company (SLOVENSKE ELEKTRARNE-SE) for upgrading the Units 1 and 2 of Mochovce Nuclear Power Plant according to the western safety standards. This is the first important project involving west-european companies in the modernisation of Russian type of pressurized water reactor (VVER 440/213). The consortium will cooperate with other partners involved in the project: Slovak, Czech and Russian. The financing of the project will be provided mainly form Slovak and Czech sources. The safety upgrading will be financed through French and German buyer credits. French company Electricite de France (EDF) will be the consultant for SE. The safety upgrading measures have been elaborated taking into account the recommendation of Vienna International Atomic Energy Agency (IAEA) and the evaluation of the safety realised by RISKAUDIT, the common organization of German and French safety authorities (GSR and IPSN). Hence all guaranties have been taken to fulfil the western safety criteria for Nuclear Power Plant Mochovce. (author)

  3. Krsko NPP Periodic Safety Review program

    International Nuclear Information System (INIS)

    Basic, I.; Spiler, J.; Novsak, M.

    2001-01-01

    The need for conducting a Periodic Safety Review for the Krsko NPP has been clearly recognized both by the NEK and the regulator (SNSA). The PSR would be highly desirable both in the light of current trends in safety oversight practices and because of many benefits it is capable to provide. On January 11, 2001 the SNSA issued a decision requesting the Krsko NPP to prepare a program and determine a schedule for the implementation of the program for 'Periodic Safety Review of NPP Krsko'. The program, which is required to be in accordance with the IAEA safety philosophy and with the EU practice, was submitted for the approval to the SNSA by the end of March 2001. The paper summarizes Krsko NPP Periodic Safety Review Program [1] including implemented SNSA and IAEA Expert Mission comments.(author)

  4. Ignalina NPP Safety Analysis: Models and Results

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Research directions, linked to safety assessment of the Ignalina NPP, of the scientific safety analysis group are presented: Thermal-hydraulic analysis of accidents and operational transients; Thermal-hydraulic assessment of Ignalina NPP Accident Localization System and other compartments; Structural analysis of plant components, piping and other parts of Main Circulation Circuit; Assessment of RBMK-1500 reactor core and other. Models and main works carried out last year are described. (author)

  5. Regulatory aspects of NPP safety

    International Nuclear Information System (INIS)

    Stuller, J.; Brandejs, P.; Miasnikov, A.; Svab, M.

    1999-01-01

    In beginning, a history of legislative process regulating industrial utilisation of nuclear energy is given, including detailed list of decrees issued by the first regulatory body supervising Czech nuclear installations - Czechoslovak Atomic Energy Commission (CSKAE). Current status of nuclear regulations and radiation protection, especially in connection with Atomic Act (Act No 18/1997 Coll.), is described. The Atomic Act transfers into the Czech legal system a number of obligations following from the Vienna Convention on Civil Liability for Nuclear Damage and Joint Protocol relating to the Application of the Vienna and Paris Convention, to which the Czech Republic had acceded. Actual duties and competence of current nuclear regulatory body - State Office for Nuclear Safety (SUJB) - are given in detail. Execution of the State supervision of peaceful utilisation of nuclear energy and ionising radiation is laid out in several articles of the Act, which comprises: control activities of the SUJB, remedial measures, penalties. Material and human resources are sufficient for fulfilment of the basic functions for which SUJB is authorised by the law. For 1998, the SUJB allotted staff of 149, approximately 2/3 of that number are nuclear safety and radiation protection inspectors. The SUJB budget for 1998 is approximately 180 million Czech crowns (roughly 6 million US dollars). Inspection activity of SUJB is carried out in three different ways: routine inspections, planned specialised inspections, inspections as a response to a certain situation (ad-hoc inspections). Approach to the licensing of major plant upgrades and backfittings are mainly illustrated on the Temelin NPP licensing. Regulatory position and practices concerning review activities are presented. (author)

  6. Safety culture development at Daya Bay NPP

    International Nuclear Information System (INIS)

    Zhang Shanming

    2001-01-01

    From view on Organization Behavior theory, the concept, development and affecting factors of safety culture are introduced. The focuses are on the establishment, development and management practice for safety culture at Daya Bay NPP. A strong safety culture, also demonstrated, has contributed greatly to improving performance at Daya Bay

  7. Development of NPP safety regulation in Russia

    International Nuclear Information System (INIS)

    Vishnevsky, Y.G.; Gutsalov, A.T.; Bukrinsky, A.M.; Gordon, B.G.

    1999-01-01

    The presentation describes the organisation scheme of Russian safety regulatory bodies, their tasks and responsibilities. Legislative and regulatory basis of NPP safety regulations rely on the federal laws: Law on the Use of Nuclear Energy and Law on Radiation Safety of the Population. Role of international cooperation and Improvement of regulatory activities in Russia are emphasised

  8. Psychology of NPP operation safety

    International Nuclear Information System (INIS)

    Tret'yakov, V.P.

    1993-01-01

    The book is devoted to psychologic investigations into different aspects of NPP operative personnel activities. The whole set of conditions on which successful and accident-free personnel operation depends, is analysed. Based on original engineering and socio-psychologic investigations complex psychologic support for NPP personnel and a system of training and upkeep of operative personnel skills are developed. The methods proposed have undergone a practical examination and proved their efficiency. 154 refs., 12 figs., 9 tabs

  9. Safety upgrading program in NPP Mochovce

    International Nuclear Information System (INIS)

    Baumeister, P.

    1999-01-01

    EMO interest is to operate only nuclear power plants with high standards of nuclear safety. This aim EMO declare on preparation completion and commissioning of Mochovce Nuclear Power Plant. Wide co-operation of our company with International Atomic Energy Agency and west European Inst.ions and companies has been started with aim to fulfil the nuclear safety requirements for Mochovce NPP. Set of 87 safety measures was implemented at Mochovce Unit 1 and is under construction at Unit 2. Mochovce NPP approach to safety upgrading implementation is showed on chosen measures. This presentation is focused on the issues category III.(author)

  10. Safety improvements of Temelin NPP

    International Nuclear Information System (INIS)

    Vita, J.

    2000-01-01

    A detailed overview is given of the efforts made to enhance the safety level of the plant considering recommendations of a number of assessment missions. A list is presented of 10 international missions of the IAEA at the Temelin plant, covering the period 1990 to 1998. For each mission the date and objective is given, the focus of the assessment is characterized, the international participation of experts is specified, and the main conclusions of the experts is reproduced. A commented list of 60 main design changes and safety improvements is also included, as they were implemented in the wake of various safety assessments. An overview of the Temelin safety improvement programme is attached, comprising brief descriptions of 30 planned improvement items together with the time schedules. (A.K.)

  11. Safety upgrading of Bohunice V1 NPP

    International Nuclear Information System (INIS)

    2001-01-01

    This CD is multimedia presentation of programme safety upgrading of Bohunice V1 NPP. It consist of next chapters: (1) Introductory speeches; (2) Nuclear power plant WWER 440; (3) Safety improvement; (4) Bohunice Nuclear power plants subsidiary; (5) Siemens; (6) REKON; (7) VUJE Trnava, Inc. - Engineering, Design and Research Organisation; (8) Album

  12. A brief overview of Ignalina NPP safety issues

    International Nuclear Information System (INIS)

    Almenas, K.; Ushpuras, E.

    1998-01-01

    A description of the safety of Ignalina NPP in a very popular form is presented. Answers to the most frequently recurring questions concerning the Ignalina NPP are provided based on recently completed international studies. Questions are like these: can a similar accident to the one that occurred in Chernobyl take place at Ignalina NPP, does the Ignalina NPP have a containment, what are the probabilities and potential consequences of accidents, etc. The brochure contains a short description of Ignalina NPP safety improvement programs

  13. Distance determination of NPP and oil reservoir on enhanced oil recovery based on heat loss and safety in view point

    International Nuclear Information System (INIS)

    Erlan Dewita; Dedy Priambodo; Sudi Ariyanto

    2013-01-01

    EOR is a method used to increasing oil recovery by injecting material or other to the reservoir. There are 3 EOR technique have been used in the world, namely thermal injection, chemical injection dan Miscible. Thermal injection method is the method most widely used in the world, however, one drawback is the loss of heat during steam distribution to the injection wells. In Indonesia, EOR application has been successfully done in the field of Duri, Chevron uses steam injection method, but still use petroleum as a fuel for steam production. In order to save oil reserves, it was done the introduction of co-generation nuclear power plants to supply some of the heat of nuclear power plants for EOR processes. In cogeneration nuclear power plant, the safety aspect is main priority. The purpose of the study was to evaluate the distance NPP with oil wells by considering heat loss and safety aspects. The method of study and calculations done using Tempo Cycle program. The study results showed that in the distance of 400 meter as exclusion zone of PBMR reactor, with pipe insulation thickness 1 in, the amount of heat loss of 277, 883 kw, while in pipe isolation thickness 2 in, amount of heat loss became 162,634 kw and with isolation thickness 3 in, amount of heat loss 120,767 kw., heat loss can be overcome and provide insulation pipes and improve the quality of saturated steam into superheated. (author)

  14. Design safety improvements of Kozloduy NPP

    International Nuclear Information System (INIS)

    Hinovski, I.

    1999-01-01

    Design safety improvements of Kozloduy NPP, discussed in detail, are concerned with: primary circuit integrity; reactor pressure vessel integrity; primary coolant piping integrity; primary coolant overpressure protection; leak before break status; design basis accidents and transients; severe accident analysis; improvements of safety and support systems; containment/confinement leak tightness and strength; seismic safety improvements; WWER-1000 control rod insertion; upgrading and modernization of Units 5 and 6; Year 2000 problem

  15. Safety Analysis Report for Ignalina NPP

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1997-01-01

    In December 1994 an agreement was signed between the European Bank for Reconstruction and Development and the Republic of Lithuania for the grant of 32.86 MECU for the safety Improvement at Ignalina NPP. One of the conditions for the provision of the grant, was a requirement for an in-depth analysis of the safety level at Ignalina NPP in the scope and according to the standards acceptable for a western nuclear power plant, and to publish a Safety Analysis Report (SAR). The report should investigate and analyze any factor that could limit a safe operation of the plant, and provide recommendations for actual safety improvements. According to the agreement, Lithuania had to finalize the SAR until 31 December, 1995. The bank has also organized and financed investigation of safety at Ignalina NPP and preparation of the SAR. EBRD made an agreement with Sweden's Vattenfall, which subcontracted well-known companies from Canada, USA, Germany, etc., and also the Russian Research and Development Institute of Power Engineering (NIKIET), reactor designer of Ignalina NPP. The SAR is a very comprehensive document and contains about 8000 pages of text, diagrams and tables. The main findings of the SAR are provided in the article. A large number of discrepancies with modern rules and western practices was detected, but they were not proved to be serious enough to require reactors shutdown. Based on the recommendations of the SAR Ignalina NPP has worked out Safety Improvement Program No. 2 (SIP-2), which is planned for three years and will cost 486 MLT. (author)

  16. NPP Temelin. Status of safety improvements

    International Nuclear Information System (INIS)

    1999-01-01

    The WWER-1000 Temelin NPP under construction has been subjected as other NPPs of the same type to numerous project reviews resulting in quite a number of recommendations for design changes. Results of the IAEA mission to review the resolution of WWER-1000 safety issues at Temelin NPP are cited in this paper. The main conclusions emphasize that a combination of eastern and western technology and practices led to safety improvements in comparison with the international practices. Plant managers are clearly committed to implementation of operational programs which are consistent with effective western operational safety practices. Considerable effort remains to bring planned programs to successful implementation, in particular in meeting the need to foster strong safety culture among all personnel

  17. NPP Temelin safety analysis reports and PSA status

    International Nuclear Information System (INIS)

    Mlady, O.

    1999-01-01

    To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

  18. Safety parameter display system for Kalinin NPP

    International Nuclear Information System (INIS)

    Andreev, V.I.; Videneev, E.N.; Tissot, J.C.; Joonekindt, D.; Davidenko, N.N.; Shaftan, G.I.; Dounaev, V.G.; Neboyan, V.T.

    1995-01-01

    The paper discusses the safety parameter display system (SPDS), which is being designed for Kalinin NPP. The assessment of the safety status of the plant is done by the continuous monitoring of six critical safety functions and the corresponding status trees. Besides, a number of additional functions are realized within the scope of KlnNPP, aimed at providing the operator and the safety engineer in the main control room with more detailed information in accidental situation as well as during the normal operation. In particular, these functions are: archiving, data logs and alarm handling, safety actions monitoring, mnemonic diagrams indicating the state of main technological equipment and basic plant parameters, reference data, etc. As compared with the traditional scope of functions of this kind of systems, the functionality of KlnNPP SPDS is significantly expanded due to the inclusion in it the operator support function ''computerized procedures''. The basic SPDS implementation platform is ADACS of SEMA GROUP design. The system architecture includes two workstations in the main control room: one is for reactor operator and the other one for safety engineer. Every station has two CRT screens which ensures computerized procedures implementation and provides for extra services for the operator. Also, the information from the SPDS is transmitted to the local crisis center and to the crisis center of the State utility organization concern ''Rosenergoatom''. (author). 3 refs, 6 figs, 1 tab

  19. Safety Culture Survey in Krsko NPP

    International Nuclear Information System (INIS)

    Strucic, M.; Bilic Zadric, T.

    2008-01-01

    The high level of nuclear safety, stability and competitiveness of electricity production, and public acceptability are the main objectives of Krsko Nuclear Power Plant. This is achievable only in environment where strong Safety Culture is taking dominant place in the way how employees communicate, perform tasks, share their ideas and attitudes, and demonstrate their concern in all aspects of work and coexistence. To achieve these objectives, behaviour of all employees as well as specific ethical values must become more transparent and that must arise from the heart of organization. Continuous ongoing and periodic self assessments of Safety Culture in Krsko NPP present major tools in implementation process of this approach. Benefits from Periodic interdisciplinary focused self assessment approach, which main intention is finding the strengths and potential areas for improvements, was used second time to assess the area of Safety Culture in Krsko NPP. Main objectives of self assessment, performed in 2006, were to increase the awareness of the present culture, to serve as a basis for improvement and to keep track of the effects of change or improvement over a longer period of time. For the purpose of effective self assessment, extensive questionnaire was used to obtain information that is representative for whole organization. Wide range of questions was chosen to cover five major characteristics of safety culture: Accountability for safety is clear, Safety is integrated into all activities, Safety culture is learning-driven, Leadership for safety is clear and Safety is a clearly recognized value. 484 Krsko NPP employees and 96 contractors were participated in survey. 70-question survey provided information that was quantified and results compared between groups. Anonymity of participant, as well as their willingness to contribute in this assessment implicates the high level of their openness in answering the questions. High number of participant made analysis of

  20. Sitting Safety Aspects of Second Romanian NPP

    International Nuclear Information System (INIS)

    Mauna, T.

    2010-01-01

    The first Romanian NPP CANDU 6 type reactor gone to erection in 1980 on Cernavoda site planned to have 5 units like the Wolsong applied design project for nuclear island. For the BOP parts the ASALDO-GE project was applied with the careful about the interface connection NSP requirements. The new NPP sitting studies began from 1982 in a serious manner as first part on Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. For develop the all package of the studies in concordance with the first IAEA Safety Standards recommendations. Till the 1982 the first mission of design and research multi-branch of specialists team was to adapt the NPP Cernavoda project having a open water cooling circuit to the new parameters of close water cooling circuit. But the team was looking at the other type of NPP for sitting. Also in the same time was studied the possibility of NSP foundation on hard less or soft soil foundation strata in connection with safety aspects. The close circuit of cooling water means others parameters of systems and need very large cooling towers. Also must be reconsidering the safety systems design and performance as new solution. In the south of Transylvania historical region in Romania the Olt River run from west to east having medium multi annual flow around 70 m3/s. The Olt River has a chain of small hydropower in operation and other planned. From geological and geophysical points of view two main faults, along the Olt river valley, one of this having seismically small activities was detected. Site region geotechnical studies show small quantity underground natural gas, salt and peat. The initial nuclear program has imposed 4 NPP units site near Olt River. Taking into account the orogenesis, water cooling needs and other local feature can't be built more than two NPP units on a site. This paper tries to reconsider the old analysis from the last IAEA Safety Standards point of view taking into account the new

  1. Mochovce NPP safety improvement and completion

    International Nuclear Information System (INIS)

    1997-01-01

    6th Nuclear society information meeting dealt with the completion of the Mochovce NPP with regard to implementation of safety measures. It was aimed to next problems: I. 'Survey' presentation on the situation of the nuclear power industry in partner countries; II. Basic technical presentations; III. Presentations of operators of the other VVER 440/213 NPPs on their activities in the field of safety improvement in relation to IAEA recommendations; IV. Technical solutions of safety improvements ranked with IAEA degree 3 (Report SC 108 VVER); V: Technical solutions of selected Safety Measures ranked with IAEA degree 2 and 1 (Report SC 108 VVER)

  2. NPP Krsko Periodic Safety Review action plan

    International Nuclear Information System (INIS)

    Bilic Zabric, T.

    2006-01-01

    In the current, internationally accepted, safety philosophy Periodic Safety Reviews (PSRs) are comprehensive reviews aimed at the verification that an operating NPP remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain an acceptable level of safety. These reviews are complementary to the routine and special safety reviews. They are long time-scale reviews intended to deal with the cumulative effects of plant ageing, modifications, operating experience and technical developments, which are not so easily comprehended over the shorter time-scale routine of safety reviews. The review was completed in 2005 and the next period will see the implementation of the action plan including some plant upgrades. The action plan lists issues that should be implemented at NPP Krsko together with associated milestones. The milestones were assumed based on best estimate resource availability and their ends can be potentially floated. In some cases, multiple corrective measures may be postulated to provide resolution for a given safety issue. The Slovenian Nuclear Safety Administration by decree approved the first periodic safety review and the implementation plan of activities arising from it. The entire implementation plan must be carried out by 15 October 2010. Report on the second periodic safety review must be submitted by the NEK not later than 15 December 2013. (author)

  3. Preparation of NPP Dukovany periodic safety review

    International Nuclear Information System (INIS)

    Dubsky, L.; Vymazal, P.

    2004-01-01

    Dukovany NPP in Czech Republic performs a periodic safety review for the second time after approximately 20 years of operation. The history of the Safety Report and its transformation into an internationally accepted form complying with IAEA standards is described. The deterministic and probabilistic assessment of the plant's safety-related design and state is applied to determine whether and to what extend the relevant protective goals are fulfilled by the existing plant design. A description of the step-by-step process is presented together with the creation of methods and criteria for PSR evaluation prepared by Nuclear Research Institute Rez

  4. Safety of NPP with WWER-440 and WWER-1000 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Balabanov, E [Energoproekt, Sofia (Bulgaria); Gledachev, J; Angelov, D [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1996-12-31

    The WWER-440 and WWER-1000 reactors used at the Kozloduy NPP have been analyzed in terms of safety. There are currently 4 reactors WWER-440/230 and 2 reactors WWER-1000/320. The former do not comply completely with the modern safety requirements due to the regulations acted in the sixties when they have been designed. The main features of these reactors are: low power density in the core; three levels of reactor control and protection; six primary loops; horizontal steam generators; two turbines; large number of cross-unit connections. The low thermal density in the core, the low specific thermal loading in the rods and the large coolant inventory enhance the safety, while the major deficiencies are identified as follows: insufficient capabilities for emergency core cooling; low diversification and physical separation of the safety systems; old fashioned control systems; inadequate fire protection; lack of full containment. It is pointed out that several design and operation actions have been completed in the Kozloduy NPP in order to enhance their safety. The WWER-1000 units are 320 model and feature a high safety level, complying completely with OPB-82 regulations and with all current international safety standards. 3 refs., 7 figs., 1 tab.

  5. Safety of NPP with WWER-440 and WWER-1000 reactors

    International Nuclear Information System (INIS)

    Balabanov, E.; Gledachev, J.; Angelov, D.

    1995-01-01

    The WWER-440 and WWER-1000 reactors used at the Kozloduy NPP have been analyzed in terms of safety. There are currently 4 reactors WWER-440/230 and 2 reactors WWER-1000/320. The former do not comply completely with the modern safety requirements due to the regulations acted in the sixties when they have been designed. The main features of these reactors are: low power density in the core; three levels of reactor control and protection; six primary loops; horizontal steam generators; two turbines; large number of cross-unit connections. The low thermal density in the core, the low specific thermal loading in the rods and the large coolant inventory enhance the safety, while the major deficiencies are identified as follows: insufficient capabilities for emergency core cooling; low diversification and physical separation of the safety systems; old fashioned control systems; inadequate fire protection; lack of full containment. It is pointed out that several design and operation actions have been completed in the Kozloduy NPP in order to enhance their safety. The WWER-1000 units are 320 model and feature a high safety level, complying completely with OPB-82 regulations and with all current international safety standards. 3 refs., 7 figs., 1 tab

  6. External flood probabilistic safety analysis of a coastal NPP

    International Nuclear Information System (INIS)

    Pisharady, Ajai S.; Chakraborty, M.K.; Acharya, Sourav; Roshan, A.D.; Bishnoi, L.R.

    2015-01-01

    External events pose a definitive challenge to safety of NPP, solely due to their ability to induce common cause failures. Flooding incidents at Le Blayais NPP, France, Fort Calhoun NPP, USA and Fukushima Daiichi have pointed to the importance of external flooding as an important contributor to NPP risk. A methodology developed for external flood PSA of a coastal NPP vulnerable to flooding due to tsunami, cyclonic storm and intense local precipitation is presented in this paper. Different tasks for EFPSA has been identified along with general approach for completing each task

  7. Conference: Nuclear Safety at the Ignalina NPP - Achievements and Challenges

    International Nuclear Information System (INIS)

    1999-01-01

    Brief description of conference which took place in Vilnius on 8-9 April 1999 is presented. The aim of the international conference is to review the work carried out, to evaluate its influence on the nuclear safety, to get acquainted with the safety system of the Ignalina NPP. On December 31, the Ignalina NPP celebrated its 15 year anniversary. During that period the Ignalina NPP has produced more than 182 billion kWh of electricity. Nuclear power is not only immense economic benefit but also a syndrome of fear of an atomic bomb and the Chernobyl accident that has formed over the years. Numerous comprehensive international studies have been performed at the Ignalina NPP. After the Chernobyl accident many steps have been taken to improve the safety of the Ignalina NPP. The introduced changes guaranteed that the void coefficient remained negative in all possible circumstances. The Ignalina NPP is intensively preparing for licensing of unit 1

  8. Probabilistic assessment of NPP safety under aircraft impact

    International Nuclear Information System (INIS)

    Birbraer, A.N.; Roleder, A.J.; Arhipov, S.B.

    1999-01-01

    Methodology of probabilistic assessment of NPP safety under aircraft impact is described below. The assessment is made taking into account not only the fact of aircraft fall onto the NPP building, but another casual parameters too, namely an aircraft class, velocity and mass, as well as point and angle of its impact with the building structure. This analysis can permit to justify the decrease of the required structure strength and dynamic loads on the NPP equipment. It can also be especially useful when assessing the safety of existing NPP. (author)

  9. A reliability evaluation method for NPP safety DCS application software

    International Nuclear Information System (INIS)

    Li Yunjian; Zhang Lei; Liu Yuan

    2014-01-01

    In the field of nuclear power plant (NPP) digital i and c application, reliability evaluation for safety DCS application software is a key obstacle to be removed. In order to quantitatively evaluate reliability of NPP safety DCS application software, this paper propose a reliability evaluating method based on software development life cycle every stage's v and v defects density characteristics, by which the operating reliability level of the software can be predicted before its delivery, and helps to improve the reliability of NPP safety important software. (authors)

  10. Safety factors for neutron fluences in NPP safety assessment

    International Nuclear Information System (INIS)

    Demekhin, V.L.; Bukanov, V.N.; Il'kovich, V.V.; Pugach, A.M.

    2016-01-01

    In accordance with global practice and a number of existing regulations, the use of conservative approach is required for the calculations related to nuclear safety assessment of NPP. It implies the need to consider the determination of neutron fluence errors that is rather complicated. It is proposed to carry out the consideration by the way of multiplying the neutron fluences obtained with transport calculations by safety factors. The safety factor values are calculated by the developed technique based on the theory of errors, features of the neutron transport calculation code and the results obtained with the code. It is shown that the safety factor value is equal 1.18 with the confidence level of not less than 0.95 for the majority of VVER-1000 reactor places where neutron fluences are determined by MCPV code, and its maximum value is 1.25

  11. The psychological background about human error and safety in NPP

    International Nuclear Information System (INIS)

    Zhang Li

    1992-01-01

    A human error is one of the factors which cause an accident in NPP. The in-situ psychological background plays an important role in inducing it. The author analyzes the structure of one's psychological background when one is at work, and gives a few examples of typical psychological background resulting in human errors. Finally it points out that the fundamental way to eliminate the unfavourable psychological background of safety production is to establish the safety culture in NPP along with its characteristics

  12. Nuclear Power Plant (NPP) safety in Brazil

    International Nuclear Information System (INIS)

    Lederman, L.

    1980-01-01

    The multidisciplinary aspects of the activities involved in the nuclear power plant (NPP) licensing, are presented. The activities of CNEN's technical staff in the licensing of Angra-1 and Angra-2 power plants are shown. (E.G.) [pt

  13. Qinshan NPP large break LOCA safety analysis

    International Nuclear Information System (INIS)

    Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

    1997-01-01

    Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

  14. The Bohunice NPP V-1 units nuclear safety upgrading

    International Nuclear Information System (INIS)

    Mlcuch, M.

    2000-01-01

    Safety upgrading and operational reliability improvement was carried out by the Bohunice NPP V-1 staff continuously since the plant commissioning. By now, more than 1200 minor or major modifications have been implemented, either by the NPP maintenance staff or by the contractors. Based on findings of safety assessment missions invited by Bohunice NPP in 1990 - 1991, the Czecho-slovak Nuclear Regulatory Authority (CSKAE) issued the decision No. 5/91 of 81 safety upgrading measures to be taken in different areas. These improvements are referred to as the 'Small Reconstruction of the Bohunice V-1 NPP'. Realization of measures during Small reconstruction of the Bohunice NPP V-1 became a power plant, which further operation is acceptable from safety point of view, but it is also necessary further safety improvement. During the period of the Small Reconstruction the development of a Safety Report for the Gradual reconstruction has been completed. Based on this report the SR Nuclear Regulatory Authority issued the Decision No. 1/94, in which requires 59 upgrading measures in different areas to be addressed. The development of Basic Engineering of the Gradual Reconstruction has been contracted to the Siemens AG. Implementation of safety measures are provided through contract with the consortium REKON (which consists of Siemens AG company and Nuclear Power Plants Research Institute Trnava) and other Czech, Russian and Slovak companies. The Gradual Reconstruction of Bohunice NPP V-1 will be finished in 2000. By implementation of the measures carried out during Gradual Reconstruction achievement of an internationally acceptable nuclear safety level will be reached. (author)

  15. NPP safety and personnel training. XII International conference. Abstracts

    International Nuclear Information System (INIS)

    2011-01-01

    The 12th International conference NPP Safety and Personnel Training took place in Obninsk, October 4-7, 2011. The issues of nuclear technologies safety are considered.The problems of life-cycle management of nuclear facilities are discussed. The criteria of assessment of physical protection systems of nuclear facilities are presented [ru

  16. Regulatory review of NPP Krsko Periodic Safety Review

    International Nuclear Information System (INIS)

    Lovincic, D.; Muehleisen, A.; Persic, A.

    2004-01-01

    At the request of the Slovenian Nuclear Safety Administration (SNSA), Krsko NPP prepared a Periodic Safety Review (PSR) program in January 2001. This is the first PSR of NPP Krsko, the only nuclear power plant in Slovenia. The program was reviewed by the IAEA mission in May 2001 and approved by SNSA in July 2001. The program is made in accordance with the IAEA Safety Guide 'Periodic Safety Review of Operational Nuclear Power Plants' No. 50-SG-012 and with European practice. It contains a systematic review of operation of the NPP Krsko, including the review of the changes as a result of the modernization of the facility. The main tasks of PSR are review of plant status for each safety factor, development of aging and life cycle management program, review of seismic design and PSHA analysis and update of regulatory compliance program. The prioritization process of findings and action plan are also important tasks of PSR. The basic safety factors of the PSR review are: Operational Experience, Safety Assessment and Analyses, Equipment Qualification and Ageing Management, Safety Culture, Emergency Planing, Environmental Impact and Radioactive Waste, Compliance with license requirements and Prioritization. It had been agreed that SNSA will have reviewed all PSR reports generated during the PSR process. At the end of 2003 the PSR Summary Report with selected recommendations for action plan was completed and delivered to SNSA for review. The paper presents regulatory review of NPP Krsko PSR with emphasis on the evaluation of the PSR issues ranking process. (author)

  17. Cernavoda NPP: Training for safety and reliability

    International Nuclear Information System (INIS)

    Postolache, Laura Lia

    2001-01-01

    The safe and reliable operation of NPP require successful integration of plant and system design (1), programmes and procedures (2) and qualified human resources (3). Of these three components, station personnel and management have capability to influence and improve programmes and competence of qualified personnel. Qualifying personnel includes selection, training and evaluation that meet the established performance standards. Training, therefore prepares people to achieve such competence. The critical role of operations personnel has been rightly emphasized by every country with a nuclear power programme. So far as operation team is concerned, they have to work, on the one hand with exacting safety rules and at the same time, they have to do the right thing at all times. In essence, they have to be prepared for new, emergency situations as well as for routine work. The plant operation in the Control Room is essentially a man - machine interaction and a safe and reliable operation requires them to take high quality decisions even under stressful conditions. Here lies therefore the need for high competent and licensed operations engineers who will ensure operation within the operating license of the station under the all conditions. The development of a long-term comprehensive training for Operation Staff is a requirement. The program addresses the qualification requirements of the various nuclear positions on shift, the outline content of the required training programs and the evaluation per the Systematic Approach to Training (SAT). A nuclear operator's training begins the moment he/she enters the station. It takes four to six years to develop the skills required to demonstrate that the candidate is an appropriate choice for the position. Then there's a further about two years of intense training at the Training Center on a simulator. After successful completion of the program, the candidate is authorized by the CNCAN (National Commission for Control of Nuclear

  18. A risk informed safety classification for a Nordic NPP

    International Nuclear Information System (INIS)

    Jaenkaelae, K.

    2002-01-01

    The report describes a study to develop a safety classification proposal or classi- fication recommendations based on risks for selected equipment of a nuclear power plant. The application plant in this work is Loviisa NPP unit 1. The safety classification proposals are to be considered as an exercise in this pilot study and do not necessarily represent final proposals in a real situation. Comparisons to original safety classifications and technical specifications were made. The study concludes that it is possible to change safety classes or safety signifi- cances as considered in technical specifications and in in-service-inspections into both directions without endangering the safety or even by improving the safety. (au)

  19. Engineering approach to relative quantitative assessment of safety culture and related social issues in NPP operation

    International Nuclear Information System (INIS)

    Sivokon, V.; Gladyshev, M.; Malkin, S.

    2005-01-01

    The report is devoted to presentation of engineering approach and software tool developed for Safety Culture (SC) assessment as well as to the results of their implementation at Smolensk NPP. The engineering approach is logic evolution of the IAEA ASSET method broadly used at European NPPs in 90-s. It was implemented at Russian and other plants including Olkiluoto NPP in Finland. The approach allows relative quantitative assessing and trending the aspects of SC by the analysis of evens features and causes, calculation and trending corresponding indicators. At the same time plant's operational performances and related social issues, including efficiency of plant operation and personnel reliability, can be monitored. With the help of developed tool the joint team combined from personnel of Smolensk NPP and RRC 'Kurchatov Institute' ('KI') issued the SC self-assessment report, which identifies: families of recurrent events, main safety and operational problems ; their trends and importance to SC and plant efficiency; recommendations to enhance SC and operational performance

  20. NPP Krsko core calculations to improve operational safety

    International Nuclear Information System (INIS)

    Ivekovic, I.; Grgic, D.; Nemec, T.

    2007-01-01

    Calculation tools and methodology used to perform independent calculations of cumulative influence of different changes related to fuel and core operation of NPP Krsko were described. Some examples of steady state and transient results are used to illustrate potential improvements to understanding and reviewing plant safety. (author)

  1. Seismic safety of building structures of NPP Kozloduy III

    International Nuclear Information System (INIS)

    Varbanov, G.I.; Kostov, M.K.; Stefanov, D.D.; Kaneva, A.D.

    2005-01-01

    In the proposed paper is presented a general summary of the analyses carried out to evaluate the dynamic behavior and to assess the seismic safety of some safety related building structures of NPP Kozloduy. The design seismic loads for the site of Kozloduy NPP has been reevaluated and increased during and after the construction of investigated Units 5 and 6. Deterministic and probabilistic approaches are applied to assess the seismic vulnerability of the investigated structures, taking into account the newly defined seismic excitations. The presented results show sufficient seismic safety for the studied critical structures and good efficiency of the seismic upgrading. The applicability of the investigated structures at sites with some higher seismic activities is discussed. The presented study is dealing mainly with the civil structures of the Reactor building, Turbine hall, Diesel Generator Station and Water Intake Structure. (authors)

  2. Safety reassessment of the Paks NPP (the AGNES project)

    Energy Technology Data Exchange (ETDEWEB)

    Gado, J [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics; Bajsz, J; Cserhati, A; Elter, J [Paksi Atomeroemue Vallalat, Paks (Hungary); Hollo, E [Energiagazdalkodasi Intezet, Budapest (Hungary); Kovacs, K [EROTERV Engineering and Contractor Co (Hungary); Maroti, L [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics; Miko, S [Paksi Atomeroemue Vallalat, Paks (Hungary); Techy, Z [Energiagazdalkodasi Intezet, Budapest (Hungary); Vidovszky, I [Hungarian Academy of Sciences, Budapest (Hungary). Central Research Inst. for Physics

    1996-12-31

    The reassessment of the Paks NPP safety according to internationally recognized criteria of the Advanced General and New Evaluation of Safety (AGNES) project is outlined. The Paks NPP consists of four WWER-440/V-213 units. The following groups of analysis have been performed: system analysis and description; analysis of design basis accidents; severe accidents analysis; level 1 probabilistic safety analysis. Postulated accidents (PA) and Anticipated Operational Occurrences (AOO) are estimated in detail for the following initiating events: increase/decrease in secondary heat removal; decrease in primary coolant inventory; increase/decrease of reactor coolant inventory; reactivity and power distribution anomalies; analysis of transients with the failure of reactor scram (ATWS); pressurized thermal shock analyses. Severe accident analysis was made for the accidents on in-vessel phase and containment phase, for radioactive release and for accident management.

  3. Safety reassessment of the Paks NPP (the AGNES project)

    International Nuclear Information System (INIS)

    Gado, J.; Hollo, E.; Kovacs, K.; Maroti, L.; Techy, Z.; Vidovszky, I.

    1995-01-01

    The reassessment of the Paks NPP safety according to internationally recognized criteria of the Advanced General and New Evaluation of Safety (AGNES) project is outlined. The Paks NPP consists of four WWER-440/V-213 units. The following groups of analysis have been performed: system analysis and description; analysis of design basis accidents; severe accidents analysis; level 1 probabilistic safety analysis. Postulated accidents (PA) and Anticipated Operational Occurrences (AOO) are estimated in detail for the following initiating events: increase/decrease in secondary heat removal; decrease in primary coolant inventory; increase/decrease of reactor coolant inventory; reactivity and power distribution anomalies; analysis of transients with the failure of reactor scram (ATWS); pressurized thermal shock analyses. Severe accident analysis was made for the accidents on in-vessel phase and containment phase, for radioactive release and for accident management

  4. Enhancing operational nuclear safety

    International Nuclear Information System (INIS)

    Sengoku, Katsuhisa

    2008-01-01

    Since Chernobyl, the dictum A n accident anywhere is an accident everywhere i s a globally shared perception. The paper presents challenges to the international nuclear community: globalization, sustainable and dynamic development, secure, safe and clean energy supply, nuclear r enaissance , public concern for nuclear safety, nuclear security, and technology and management. Strong national safety infrastructures and international cooperation are required to maintain a high level of nuclear safety and security worldwide. There is an increasing number of countries thinking of going nuclear: Morocco, Indonesia, Iran, Poland, Turkey, Bangladesh, Egypt, Vietnam, Chile, Nigeria, Malaysia, Thailand, Uruguay, Tunisia, Algeria. Another serious incident will jeopardize the prospect of nuclear renaissance. Safety and security are preconditions for countries newly introducing NPP as well as for those with mature nuclear programmes. The Global Nuclear Safety Regime (GNSR) is referred to as the institutional, legal and technical framework to achieve worldwide implementation of the safety of nuclear installations. At the top of the framework is the Convention on Nuclear Safety which covers the nuclear power plants. The convention has 56 contracting parties which meet triennially where national reports are presented and subject to the review of peers. The International Atomic Energy Agency (IAEA) undertakes a programme to foster the GNSR through the establishment of IAEA safety standards and related publications. The programme provides for the application of standards for the (1) safety of nuclear installations, (2) safety of radioactive sources, (3) safe transport of radioactive material and (4) management of radioactive waste. It also provides for the security of nuclear installations, nuclear material and radioactive material. The safety standards hierarchy is as follows: safety fundamental, safety requirements and safety guides. The safety fundamentals are the bases for IAEA

  5. Determination of Safety Performance Grade of NPP Using Integrated Safety Performance Assessment (ISPA) Program

    International Nuclear Information System (INIS)

    Chung, Dae Wook

    2011-01-01

    Since the beginning of 2000, the safety regulation of nuclear power plant (NPP) has been challenged to be conducted more reasonable, effective and efficient way using risk and performance information. In the United States, USNRC established Reactor Oversight Process (ROP) in 2000 for improving the effectiveness of safety regulation of operating NPPs. The main idea of ROP is to classify the NPPs into 5 categories based on the results of safety performance assessment and to conduct graded regulatory programs according to categorization, which might be interpreted as 'Graded Regulation'. However, the classification of safety performance categories is highly comprehensive and sensitive process so that safety performance assessment program should be prepared in integrated, objective and quantitative manner. Furthermore, the results of assessment should characterize and categorize the actual level of safety performance of specific NPP, integrating all the substantial elements for assessing the safety performance. In consideration of particular regulatory environment in Korea, the integrated safety performance assessment (ISPA) program is being under development for the use in the determination of safety performance grade (SPG) of a NPP. The ISPA program consists of 6 individual assessment programs (4 quantitative and 2 qualitative) which cover the overall safety performance of NPP. Some of the assessment programs which are already implemented are used directly or modified for incorporating risk aspects. The others which are not existing regulatory programs are newly developed. Eventually, all the assessment results from individual assessment programs are produced and integrated to determine the safety performance grade of a specific NPP

  6. Safety upgrade at the Leningrad NPP

    International Nuclear Information System (INIS)

    Eperin, A.P.

    1996-01-01

    The LNPP was developed according to the standards of early 70's but, at the same time, during the whole period of operation, the Plant equipment, technological, automatic and control and protection systems were upgraded with regard to changing safety and reliability requirements. Main steps taken during the backfitting stage to improve the reliability and safety of LNPP equipment and systems are discussed

  7. Safety analysis of Ignalina NPP during shutdown conditions

    International Nuclear Information System (INIS)

    Kaliatka, A.; Uspuras, E.

    2000-01-01

    The accident analysis for the Ignalina NPP with RBMK-1500 reactors at normal operating conditions and at minimum controlled power level (during startup of the reactor) has been performed in the frame of the project I n-Depth Safety Assessment of the Ignalina NPP , which was completed in 1996. However, the plant conditions during the reactor shutdown differ from conditions during reactor operation at full power (equipment status in protection systems, set points for actuation of safety and protection systems, etc.). Results of RELAP5 simulation of two worst initiating events during reactor shutdown - Pressure Header rupture in case of steam reactor cooldown as well as Pressure Header rupture in case of water reactor cooldown are discussed in the paper. Results of analysis shown that reactor are reliably cooled in both cases. Further analysis for all range of initial events during reactor shutdown and at shutdown conditions is recommended. (author)

  8. International Conference of Ukrainian Nuclear Society ''NPP's safety and protection''(annotations)

    International Nuclear Information System (INIS)

    Barbashev, S.V.

    1997-01-01

    The abstracts of reports submitted to the Conference include: - New developments of the safe nuclear installations; - NPP ecological safety; - Methods of personnel and population protection; - Waste management safety (at transportation, processing and storage); - Spent nuclear fuel management; - NPP life extension and decommissioning; - Public opinion as an element of NPP safety; - Training of personnel, scientific support and safety culture; - Forecasting of nuclear power and industry safe development; - Development of international cooperation in nuclear power

  9. Determination of the NPP Krsko reactor core safety limits using the COBRA-III-C code

    International Nuclear Information System (INIS)

    Lajtman, S.; Feretic, D.; Debrecin, N.

    1989-01-01

    This paper presents the NPP Krsko reactor core safety limits determined by the COBRA-III-C code, along with the methodology used. The reactor core safety limits determination is a part of reactor protection limits procedure. The results obtained were compared to safety limits presented in NPP Krsko FSAR. The COBRA-III-C NPP Krsko design core steady state thermal hydraulics calculation, used as the basis for the safety limits calculation, is presented as well. (author)

  10. Regulatory aspects of NPP safety in Lithuania

    International Nuclear Information System (INIS)

    Kutas, S.

    1999-01-01

    The functions of safety and control of nuclear facilities as well as supervision of accounting nuclear materials shall be performed by the State Nuclear Power safety Inspectorate of the Republic of Lithuania (VATESI). The supervision and control of nuclear facilities shall also be carries out within the framework of their respective competence, by other state institutions as well as by local authorities in the territories of their jurisdiction. Licenses for the activities referred to design, construction and reconstruction of nuclear facilities shall be issued by VATESI with the approval of the Ministry of Health, Ministry of Environment a local authority whose territory is within the sanitary protection zone of nuclear facility. An urgent need is recognised for support on integrated international assistance program for improvement of Ignalina safety program. A detailed list of nuclear regulation acts is included in the presentation

  11. VVER 1000-NPP Temelin safety upgrading

    International Nuclear Information System (INIS)

    Fleischhans, J.; Ubra, O.

    1995-01-01

    A modernisation program upgrading Temelin plant to meet internationally adopted standard has been implemented during plant design and construction phases. The initial Czech-Russian design (primary system was of Russian design, secondary system was of Czech design) has been extensively modified and adapted to present western safety criteria and operational requirements. The goals are to achieve a high level of safety, reliability, availability and load-following ability. The load-following ability and response to grid frequency changes are very important for the Czech Republic, since the nuclear capacity represents a high proportion of the overall electrical system there. On the basis of IAEA OSART missions and Halliburton NUS audit results and in compliance with recommendations of The State Office for Nuclear Safety, Czech Power Company and Czech scientists and researchers a modernisation program project for Temelin has been carried out. It includes three main groups of VVER1000 MW unit innovations: - Modernization and upgrading of the safety and control systems. - Fuel replacement and modification of the reactor core. - Innovation of some components of the primary and secondary systems. The tenders for instrumentation and control system, nuclear fuel, diagnostic system and radiation monitoring system were issued to the world-well known suppliers. The US company Westinghouse Electric >Corporation (WEC) was selected to submit contract for the delivery of instrumentation and control system primary side diagnostic system and for the delivery of nuclear fuel. The contract was signed in 1993

  12. Safety assessment and improvement of Ignalina NPP against downcomer ruptures outside Accident Localisation System

    International Nuclear Information System (INIS)

    Rimkevicius, S.; Urbonavicius, E.

    2002-01-01

    Accident Localisation System (ALS) of Ignalina NPP is a pressure suppression type confinement, designed to prevent the release of contaminated steam-water mixture to the environment in case of Loss-of-Coolant Accident (LOCA). One of the peculiarities of Ignalina NPP with RBMK-1500 reactors is that not all of the reactor coolant circuit is enclosed within ALS. Some part of downcomers, that connect Drum Separator (DS) and suction header of main circulation pump is located outside ALS. In case of downcomer rupture in DS compartment the discharge is not confined, but flows to the environment through the safety panels installed in the ceiling of DS compartments. Numerous safety analyses were performed to assess the safety of Ignalina NPP against downcomer break outside ALS, and results were used for different applications in order to improve the safety of the plant. This paper presents the overview of the performed analyses, recommendations raised and safety improvements made to enhance the safety level of NPP. One of the applications is to present the recommendations for safety improvement if maximal allowable pressure limits are exceeded. The calculations results demonstrate that in the case of two downcomers rupture in drum separators compartment the maximum permissible pressure in the reactor hall could be exceeded. The knock-out panels from the reactor hall to the environment were recommended and installed for reactor hall overpressure protection. The evaluation of the drainage system efficiency from DS compartments was performed. In this case the especial attention was paid to analyse the water collection and drainage system behaviour in long term after postulated breaks. The analysis results showed that the modernization of the drainage system prevents the accumulation of the released water in the compartments even in the case of two downcomer pipes ruptures, and decreases the release of radioactive fission products (FP) to the environment.(author)

  13. IAEA activities on safety aspects of NPP ageing

    International Nuclear Information System (INIS)

    Pachner, J.

    1998-01-01

    A review of IAEA activities concerned with safety aspects of nuclear power plants ageing is given for the period from 1995 to 1998 with the prospects till year 2000. Coordinated Research programs were conducted on Management Ageing of Concrete Containment Buildings; Management of Ageing of In-Containment I and C cables. TECDOCs were published on Assessment and Management of Ageing of Major NPP Components Important for Safety of CANDU, PWR and BWR NPPs. Technical Committee Meetings and Interregional training courses concerned with the same subjects were held

  14. Probabilistic safety analysis forecast for Trillo 1 NPP

    International Nuclear Information System (INIS)

    Carretero Fernandino, J.A.; Martin Alvarez, L.; gomez, F.; Cuallado, G.

    1995-01-01

    The performance of Probabilistic Safety Analyses (PSA) at Trillo 1 NPP is facing a number of challenges, unprecedented in previous PSAs carried out in Spain, due to the particular design characteristics of the plant. On account of this, it has been necessary to implemented specific approaches and methodological alternatives to perform a PSA which, while maintaining detail level and requirements in line with PSAs carried out previously in Spain, offers a solution technically adapted to the characteristics of the SIEMENS-KWU design as opposed to other Spanish reactors with a basic Westinghouse-General Electric design, which are based on standard US design. The purpose of this paper is to describe the most significant characteristics of the PSA at Trillo 1 NPP and the methodology used to date, taking into account current project progress

  15. Safety measures for NPP Mochovce - Skoda contribution

    International Nuclear Information System (INIS)

    Svec, P.

    1997-01-01

    Brief technical descriptions are given of the following issues involved in safety measures: reactor pressure vessel embrittlement and its monitoring; non-destructive testing; integrity of the steam generator primary collector; integrity of steam generator heat exchanging tubes; distribution of feedwater in the steam generator; the hazard of plugging water collectors located on the steam generator room floor; steam generator level control valves; primary circuit venting under emergency conditions; diagnostic systems installed in the primary circuit; monitoring of reactor pressure vessel head leaks; severity of containment leaks; turbine missiles; seismic design; procedures for normal operation; and operation limits and conditions. (M.D.)

  16. Optimization of maintenance periodicity of complex of NPP safety systems

    International Nuclear Information System (INIS)

    Kolykhanov, V.; Skalozubov, V.; Kovrigkin, Y.

    2006-01-01

    The analysis of the positive and negative aspects connected to maintenance of the safety systems equipment which basically is in a standby state is executed. Tests of systems provide elimination of the latent failures and raise their reliability. Poor quality of carrying out the tests can be a source of the subsequent failures. Therefore excess frequency of tests can result in reducing reliability of safety systems. The method of optimization of maintenance periodicity of the equipment taking into account factors of its reliability and restoration procedures quality is submitted. The unavailability factor is used as a criterion of optimization of maintenance periodicity. It is offered to use parameters of reliability of the equipment and each of safety systems of NPPs received at developing PSA. And it is offered to carry out the concordance of maintenance periodicity of systems within the NPP maintenance program taking into account a significance factor of the system received on the basis of the contribution of system in CDF. Basing on the submitted method the small computer code is developed. This code allows to calculate reliability factors of a separate safety system and to determine optimum maintenance periodicity of its equipment. Optimization of maintenance periodicity of a complex of safety systems is stipulated also. As an example results of optimization of maintenance periodicity at Zaporizhzhya NPP are presented. (author)

  17. New ICRP recommendations and radiation safety of an NPP

    International Nuclear Information System (INIS)

    Janzekovic, H.

    2007-01-01

    In March 2007 the fundamental radiation protection recommendations used world-widely in nuclear facilities were approved by the ICRP. Implementation of radiation safety standards in an NPP is a challenging issue related to all NPP phases from planning a site and its design to its decommissioning also because if neglected it could be very difficult if not impossible to implement improvement of radiation safety later during operation or decommissioning without a substantial cost. The standards are changing with a period of 15 years which is small regarding a prolonged lifetime of many NPPs and also foreseen lifetime of new NPPs, i.e. 60 years. The new recommendations are actually an upgrading of the ICRP 60. Among other changes new sets of wR and wT are given, as well as an update of around 50 different values related to doses. Two new concepts are also tackled i.e. terrorist attacks and protection of the environment. The influence of the new recommendations on the radiation safety of NPPs can be analysed by a selection of four renewed or new concepts: types of exposure situation, dose constraints, source-related approach and safety and security. Their implementation could lead to upgrading the radiation safety of future or existing NPPs as well as of decommissioning processes. Some of the concepts were already extensively and successfully used by designers of modifications or of new NPPs, as well as by operators. (author)

  18. Probabilistic safety assessment activities at Ignalina NPP

    International Nuclear Information System (INIS)

    Bagdonas, A.

    1999-01-01

    The Barselina Project was initiated in the summer 1991. The project was a multilateral co-operation between Lithuania, Russia and Sweden up until phase 3, and phase 4 has been performed as a bilateral between Lithuania and Sweden. The long-range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. During phase 3, from 1993 to 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. During phase 4, from 1994 to 1996, the PSA was further developed, taking into account plant changes, improved modelling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The model reflected the plant status before the outage 1996. During phase 4+, 1998 to 1999 the PSA model was upgraded taking into account the newest plant modifications. The new PSA model of CPS/AZRT was developed. Modelling was based on the Single Failure Analysis

  19. Operational and safety status of Krsko NPP

    International Nuclear Information System (INIS)

    Sirola, P.; Kavsek, D.

    1998-01-01

    Nuclear Power Plants Krsko (NEK) is producing electricity with the high level of reliability, safety and at acceptable price for 17 years. Energy is shared between both Slovenian and Croatian grid. The specifics of sharing the initial investment costs, later covering the operations costs and energy supply between Croatia and Slovenia is causing specific decision making problems about energy cost and future investments, however not influencing the plant safety, by now. NEK is continuously following the international nuclear technology practices, standards' changes and improvements and introducing them into the processes and equipment upgrades. As the member of the most important international integration, NEK is having the possibility of sharing its experience with others. Slovenian Energy Consumption and Supply Strategy is recognizing the NEK as a long term supply of energy in Slovenia being a strong decision making base for the future. According to the above mentioned Slovenian Energy Consumption and Supply Strategy the plant is obliged to keep all the radioactive waste, produced during the plant life, on site. The extensive efforts are taking place to reduce the radioactive waste production and save the area available for temporary waste deposition. The plant is licensed for the period of 40 years of commercial operation which started in 1983, so the Life Time Management is getting more and more important, including the performance tracing of the essential components, their maintenance and surveillance programs and also replacement plans of critical equipment. The major problems the NEK is confronted with at the moment are the Steam Generators which are reaching their and of life, and a very limited radioactive waste storage area. They are excerting influence on the plant availability and operations and maintenance costs. At the moment the process of Modernization is in progress, covering the Steam Generators replacement and a Plant Specific Simulators supply

  20. PSA analysis focused on Mochovce NPP safety measures evaluation from operational safety point of view

    International Nuclear Information System (INIS)

    Cillik, I.; Vrtik, L.

    2001-01-01

    Mochovce NPP consists of four reactor units of WWER 440/V213 type and it is located in the south-middle part of Slovakia. At present first unit operated and the second one under the construction finishing. As these units represent second generation of WWER reactor design, the additional safety measures (SM) were implemented to enhance operational and nuclear safety according to the recommendations of performed international audits and operational experience based on exploitation of other similar units (as Dukovany and J. Bohunice NPPs). These requirements result into a number of SMs grouped according to their purpose to reach recent international requirements on nuclear and operational safety. The paper presents the bases used for safety measures establishing including their grouping into the comprehensive tasks covering different areas of safety goals as well as structural organization of a project management of including participating companies and work performance. More, results are given regarding contribution of selected SMs to the total core damage frequency decreasing.(author)

  1. Passive components of NPP safety-related systems

    International Nuclear Information System (INIS)

    Ionaytis Romuald, R.; Bubnova Tatyana, A.

    2005-01-01

    This paper presents a new passive components with having drives: fast-response cutoff valves; modular actuators with opposite cocking pneumatic drives and actuation spring drives; voting electromagnetic valve units for control of pneumatic drives; passive initiators of actuation; visual diagnostics . All these devices have been developed and tested at mock-ups. This paper presents also the following direct-action passive safety components: modular pressure-relief safety valves; pilot safety valves with passive action; check valves with remote position indicator and after-tightening; modular inserts for limiting emergency coolant flow; vortex rectifier; critical weld fasteners; gas-liquid valves; fast-removable seal assembly; seal spring loaders; grooves for increasing hydraulic resistance. Replacement of active safety system components for passive ones improves the general reliability NPP by 1.5 or 2 orders of magnitudes. (authors)

  2. Safety assessment of the BELENE NPP

    International Nuclear Information System (INIS)

    Bruna, G.B.

    2010-01-01

    The most important outcome of the RISKAUDIT’s review of the BELENE’s ISAR was that the design of the plant is in general conformity with the applied requirements, the Bulgarian regulation and the international practice.No substantial design flaws have been identified and no major issues are found as missing. However, the ISAR should be significantly improved before issuing the construction permit. RISKAUDIT asked for elaboration of a revised version of the ISAR, in which all the recommendations, the requests for information as well the questions should be comprehensively addressed, and explicitly referenced to facilitate their assessment. The satisfactory accomplishment of this duty has been considered. The considered as a as prerequisite for the issuance of the construction permit, 2.The REV 2 of the ISAR have been issued at the beginning of this year, BNRA and RISKAUDIT signed on a new collaboration under the above-mentioned Contract, which has the following main objectives:a. Assessing the implementation of the above-mentioned recommendations, b.Verifying and integrating the answers provided to the questions,c.Performing additional in-depth analysis aimed at completing and finalizing the previous investigations as well as addressing some complementary relevant safety issues.3.The inception meeting of this Contract is planed June 17 –18 in Sofia. Details on the planning have to be defined, but, tentatively, the assessment activity should be completed by the end of current year and the accompanying in-depth analysis at the latest at the beginnings of 2011. 4.4.The issuance of those Reports is an important technical support., The support to to the process for granting the construction permit

  3. Meeting up-to-date safety requirements in the Russian NPP projects

    International Nuclear Information System (INIS)

    Tepkyan, G. O.; Yashkin, A. V.

    2014-01-01

    Safety features in Russian NPP designs are implemented by the combination of active and passive safety systems • Russian NPP designs are in compliance with up-to-date international and European safety requirements and refer to Generation III+ • Russian state-of-the-art designs have already implemented some design solutions, which take into account “post-Fukushima” requirements. Russian NPP design principles have been approved during the European discussions in spring 2012, including the IAEA extraordinary session addressed to Fukushima NPP accident

  4. Approaches to the mathematical description of NPP operational safety management and oversight

    International Nuclear Information System (INIS)

    Bilej, D.V.; Berzhanskij, S.V.

    2014-01-01

    The paper presents analysis of features related to NPP operational safety management and oversight. According to analysis results, approaches are proposed to perform mathematical description of specific processes and to develop a scale for management to the current safety level as regards NPP power generation. Proposed approaches are making experimental equations and process approach of ISO-9001 quality system

  5. Safety improvement and results of commissioning of Mochovce NPP WWER 440/213

    International Nuclear Information System (INIS)

    Lipar, M.

    1998-01-01

    Mochovce NPP is the last one of this kind and compared to its predecessors, it is characterized by several modifications which contribute to the improvement of the safety level. In addition based on Nuclear Regulatory Authority requirements and based on documents: - IAEA - Safety Issues and their ranking for NPP WWER 440/213, - IAEA - Safety Improvement of Mochovce NPP Project Review Mission, - Riskaudit - Evaluation of the Mochovce NPP Safety Improvements. Additional safety measures have been implemented before commissioning. The consortium EUCOM (FRAMATOME - SIEMENS), SKODA Praha, ENERGOPROJEKT Praha, Russian organizations and VUJE Trnava Nuclear Power Plants research institute were selected for design and implementation of the safety measures. The papers summarized, safety requirements, safety measures implemented, results of commissioning and results of safety analysis report evaluation. (author)

  6. Criticality safety analysis of the NPP Krsko storage racks

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2002-01-01

    NPP Krsko is going to increase the capacity of the spent fuel storage pool by replacement of the existing racks with high-density racks. This will be the second reracking campaign since 1983 when storage was increased from 180 to 828 storage locations. The pool capacity will increase from 828 to 1694 with partial reracking by the spring 2003. The installed capacity will be sufficient for the current design plant lifetime. Complete reracking of the spent fuel pool will additionally increase capacity to 2321 storage locations. The design, rack manufacturing and installation has been awarded to the Framatome ANP GmbH. Burnup credit methodology, which was approved by the Slovenian Nuclear Safety Administration in previous licensing of existing racks, will be again implemented in the licensing process with the recent methodology improvements. Specific steps of the criticality safety analysis and representative results are presented in the paper.(author)

  7. Perspectives on managing nuclear safety at Cernavoda NPP past, present and future

    International Nuclear Information System (INIS)

    Serban, M.

    1997-01-01

    The presentation considers the following issues: background of Romanian nuclear programme; 1990 management issues involved in Cernavoda project, nuclear safety perspectives; Cernavoda U1 operating organization today (safety related issues); good practices at Cernavoda NPP

  8. Perspectives on managing nuclear safety at Cernavoda NPP past, present and future

    Energy Technology Data Exchange (ETDEWEB)

    Serban, M [Technical Safety Directorate, Unit 1., Cernavoda Nuclear Power Plant (Romania)

    1997-12-01

    The presentation considers the following issues: background of Romanian nuclear programme; 1990 management issues involved in Cernavoda project, nuclear safety perspectives; Cernavoda U1 operating organization today (safety related issues); good practices at Cernavoda NPP.

  9. Reactor safety instrumentation of Paks NPP (experience and perspective)

    International Nuclear Information System (INIS)

    Elo, S.; Hamar, K.

    1993-01-01

    The majority of the existing control and protection systems in nuclear power plants use old analog technology and design philosophy. Maintenance and the procurement of spare parts is becoming increasingly difficult. In general there is an age degradation concern. Aging degradation in nuclear power plants must be effectively managed to avoid a loss of vital safety function, shutdown of the station, a reduced power generation, or any failure leading to expensive repair. Even with the best efforts in developing reliable and long life instrumentation and control systems for nuclear power plants it is expected that these systems for most plants will require replacements during the life of the plants. The instrumentation and control system of the nuclear power plants designed during the 70's and constructed in the 80's went out-of-date since nuclear safety is not a static concept and the digital computer technology has undergone rapid improvements during the 70's and 80's. Simultaneously the operation and the maintenance of the I ampersand C system of those plants described above becomes more and more difficult and expensive. In this context the pure quality of the former Soviet designed process instrumentation system increases the needs of upgrading this system. The author reviews the main design characteristics of the reactor safety instrumentation of the Paks NPP. Further he attempts to convey the perspective on upgrading the reactor safety instrumentation as seen by the HAEC and its Nuclear Safety Inspectorate

  10. South Ukraine NPP: Safety improvements through Plant Computer upgrade

    International Nuclear Information System (INIS)

    Brenman, O.; Chernyshov, M. A.; Denning, R. S.; Kolesov, S. A.; Balakan, H. H.; Bilyk, B. I.; Kuznetsov, V. I.; Trosman, G.

    2006-01-01

    This paper summarizes some results of the Plant Computer upgrade at the Units 2 and 3 of South Ukraine Nuclear Power Plant (NPP). A Plant Computer, which is also called the Computer Information System (CIS), is one of the key safety-related systems at VVER-1000 nuclear plants. The main function of the CIS is information support for the plant operators during normal and emergency operational modes. Before this upgrade, South Ukraine NPP operated out-of-date and obsolete systems. This upgrade project wax founded by the U.S. DOE in the framework of the International Nuclear Safety Program (INSP). The most efficient way to improve the quality and reliability of information provided to the plant operator is to upgrade the Human-System Interface (HSI), which is the Upper Level (UL) CIS. The upgrade of the CIS data-acquisition system (DAS), which is the Lower Level (LL) CIS, would have less effect on the unit safety. Generally speaking, the lifetime of the LL CIS is much higher than one of the UL CIS. Unlike Plant Computers at the Western-designed plants, the functionality of the WER-1000 CISs includes a control function (Centralized Protection Testing) and a number of the plant equipment monitoring functions, for example, Protection and Interlock Monitoring and Turbo-Generator Temperature Monitoring. The new system is consistent with a historical migration of the format by which information is presented to the operator away from the traditional graphic displays, for example, Piping and Instrument Diagrams (P and ID's), toward Integral Data displays. The cognitive approach to information presentation is currently limited by some licensing issues, but is adapted to a greater degree with each new system. The paper provides some lessons learned on the management of the international team. (authors)

  11. Risk-informed decision making during Bohunice NPP safety upgrading

    International Nuclear Information System (INIS)

    Lipar, M.; Muzikova, E.; Kubanyi, J.

    2001-01-01

    The paper summarizes some facts of risk-informed regulation developments within UJD regulatory environment. Based on national as well as international operating experience and indications resulted from PSA, Nuclear Regulatory Authority of the Slovak Republic (UJD) since its constituting in 1993 has devoted an effort to use PSA technology to support the regulatory policy in Slovakia. The PSA is considered a complement, not a substitute, to the deterministic approach. Suchlike integrated approach is used in decision making processes and the final decision on scope and priorities is based on it. The paper outlines risk insights used in the decision making process concerning Bohunice NPP safety upgrading and focuses on the role of PSA results in Gradual Reconstruction of Bohunice VI NPP. Besides, two other examples of the PSA results application to the decision making process are provided: the assessment of proposal of modifications to the main power supply diagram (incorporation of generator switches) and the assessment of licensee request for motor generator AOT (Allowable Outage Time) extension. As an example of improving support of Bohunice V-2 risk-informed operations, concept of AOT calculations and Bohunice V-2 Risk Monitor Project are briefly described. (author)

  12. Probabilistic safety assessment for Balakovo 1000 MW NPP

    International Nuclear Information System (INIS)

    Foden, R.W.

    1995-01-01

    In July 1993 the Commission of the European Communities (CEC) placed a contract with NNC Ltd (National Nuclear Corporation) for performing a Probabilistic Safety Assessment (PSA) for a 1000 MW NPP in the Russian Federation. The contract is part (Project 3.1) of the 1991 TACIS (Technical Assistance to the CIS) programme. This paper describes the objectives and scope of the Project and provides a description of the progress that has been made. For this Project, NNC is the leader of a Consortium of Western European companies that has been formed to undertake this Project and other Projects in the TACIS 91 programme. NNC therefore has overall responsibility for the coordination and management of the complete PSA Project. Other members of the Consortium involved in this Project are Empresarios Agrupados from Spain, Belgatom from Belgium and AEA-Technology from the UK. The analytical work for the Project is performed by the Russian Company Atomenergoproekt in Moscow, under contract to NNC. The official recipient institution for the results of the Project is the Russian Utility, Rosenergatom. The NPP chosen to be the subject of the Project is the Balakovo Unit 4 VVER 1000. (author)

  13. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu.N.; Gabaraev, B.A.

    2002-01-01

    Full text: The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper - erection of steam-gas toppings to the nuclear power units - is considered in the paper. Application of the steam-gas toppings permits through reducing power of ageing reactors to extend lifetime of nuclear power unit, enhance its safety and at the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project, for Russian boiling VK-50 reactor now in operation Application of the steam-gas topping permits: extend the service life of ageing VVER-440 reactor by 10...15 years; use the turbine and other NPP balance-of-plant equipment at full power; increase the efficiency of combined cycle up to 48% and more; enhance the safety of NPP operation; utilize NPP balance-of-plant equipment after reactor decommissioning; perform the cost-effective operation in maneuvering modes; increase capacity factor of the plant. The construction of pilot project on the basis of the VK-50 reactor will allow not only to demonstrate new technology but also to attain appreciable economic effect including that obtained due to using the available reserves of the NPP turbine. (author)

  14. Modernization and safety improvement project of the NPP V-2 Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Michal, V.; Losonsky, B.; Magdolen, J.

    2000-01-01

    This contribution deals with the form, present state, and results of the Nuclear Power Plants Research Institute (the Slovak acronym is VUJE - Vyskumny Ustav Jadrovych Elektrarni) participation in the NPP V-2 Jaslovske Bohunice Modernization and Safety Improvement Project. A short description of VUJE history, activity, and results is also presented as well as NPPs Jaslovske Bohunice characterization. VUJE was established in 1977 and deals with scientific and research needs of nuclear power plants, such as design, construction, commissioning and operation. The next fields of VUJE activity are, NPP reconstruction, NPP personnel training, radioactive waste management technology, and NPP decommissioning. The nuclear power plant, Jaslovske Bohunice, is situated approximately 15 km from the district town of Trnava in the southwestern region of the Slovak Republic. The construction of the first Czechoslovak NPP A-1 began on this site in 1957 .The construction of the double-unit NPP V-1 with WWER-440 (type V-230) reactor began in 1972. The first unit of NPP V-1 began operation in 1978 and the second in 1980. NPPs construction on the Bohunice site continued with NPP V-2, which has two units with WWER-440 (type V-213) reactors. Unit 1 and Unit 2 of NPP V-2 were commissioned in 1984 and 1985, respectively. Slovak electric utility Slovenske elektrarne (SE) is the owner/commissioner of NPP V-2. This NPP is responsible for more than 20% of the total electrical energy production of SE, making it an essential supporter of the Slovak economy. (authors)

  15. NPP Krsko periodic safety review. Safety assessment and analyses

    International Nuclear Information System (INIS)

    Basic, I.; Spiler, J.; Thaulez, F.

    2002-01-01

    Definition of a PSR (Periodic Safety Review) project is a comprehensive safety review of a plant after ten years of operation. The objective is a verification by means of a comprehensive review using current methods that the plant remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain plant safety. The overall goals of the NEK PSR Program are defined in compliance with the basic role of a PSR and the current practice typical for most of the countries in EU. This practice is described in the related guides and good practice documents issued by international organizations. The overall goals of the NEK PSR are formulated as follows: to demonstrate that the plant is as safe as originally intended; to evaluate the actual plant status with respect to aging and wear-out identifying any structures, systems or components that could limit the life of the plant in the foreseeable future, and to identify appropriate corrective actions, where needed; to compare current level of safety in the light of modern standards and knowledge, and to identify where improvements would be beneficial for minimizing deviations at justifiable costs. The Krsko PSR will address the following safety factors: Operational Experience, Safety Assessment, EQ and Aging Management, Safety Culture, Emergency Planning, Environmental Impact and Radioactive Waste.(author)

  16. New engineering safety factors for Loviisa NPP core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kuopanportti, Jaakko; Saarinen, Simo; Lahtinen, Tuukka; Ekstroem, Karoliina [Fortum Power and Heat Ltd., Fortum (Finland)

    2017-09-15

    In Loviisa NPP, there are two limiting thermal margins called the enthalpy rise margin and the linear heat rate margin that are monitored during normal operation. Engineering safety factors are applied in determination of both of these factors. The factors take into account the effect of various manufacturing tolerances, impact of the irradiation and simulation uncertainties on the local heat rate and on the enthalpy of the coolant. The engineering factors were re-evaluated during 2015 and the factors were approved by the Finnish radiation and nuclear safety authority in 2016. The re-evaluation was performed by considering all of the identified phenomena that affect the local heat rate or the enthalpy of the coolant. This paper summarizes the work that was performed during the re-evaluation of the engineering safety factors and presents the results for each uncertainty component. The new engineering safety factors are 1.115 for the linear heat rate and 1.100 for the enthalpy rise margin when the old factors were 1.12 and 1.16, respectively. The new factors improve the fuel economy by about 1%.

  17. Sociodrama approach for enhancing nuclear safety

    International Nuclear Information System (INIS)

    Choi, K. S.; Kim, C. B.; Ha, Y. H.

    2004-01-01

    A role playing or sociodrama has been experimentally conducted among residents from 4 NPP sites in Korea and KINS employees as a psychological approach for enhancing nuclear safety and improving public communication and public confidence in regulator in Dec. 2004. In this paper, the results were analyzed and presented and future plan and area of further study were suggested. This socio-psychological approach can be used as a new communication method for improving mutual understanding between residents and NPP operators at sites. It can be also used to solve conflicts among stakeholders and interest groups in nuclear industry

  18. Experience of Hungarian model project: 'Strengthening training for operational safety at Paks NPP'

    International Nuclear Information System (INIS)

    Kiss, I.

    1998-01-01

    Training of Operational Safety at Paks NPP is described including all the features of the project including namely: description of Paks NPP, its properties and performances; reasons for establishing Hungarian Model Project, its main goals, mentioning Hungarian and IAEA experts involved in the Project, its organization, operation, budget, current status together with its short term and long term impact

  19. Assessment of NPP safety taking into account seismic and engineering-geological factors

    International Nuclear Information System (INIS)

    Yakovlev, E.A.

    1990-01-01

    Consideration is given to the problem of probabilistic analysis of NPP safety with account of risk of destructive effect of earthquakes and the danger of accidental geological processes (diapirism, karst etc.) under NPP operation. It is shown that account of seismic and engineering-geological (engineering-seismological) risk factors in probabilistic analysis of safety enables to perform anticipatory analysis of behaviour of principle plant objects and to improve safety of their operation by revealing the most unstable elements of geotechnical system forming the main contribution to the total NPP risk

  20. On the increasing of NPP safety operation by the way of improving the quality of mounting works

    International Nuclear Information System (INIS)

    Girnis, V.V.

    1987-01-01

    The problem on the necessity of improving the quality of pipeline mounting to provide NPP operation safety is considered. Ways for improving the quality of NPP production equipment and pipelines welding and fit-up are given

  1. Safety analyses of the electrical systems on VVER NPP

    International Nuclear Information System (INIS)

    Andel, J.

    2004-01-01

    Energoprojekt Praha has been the main entity responsible for the section on 'Electrical Systems' in the safety reports of the Temelin, Dukovany and Mochovce nuclear power plants. The section comprises 2 main chapters, viz. Offsite Power System (issues of electrical energy production in main generators and the link to the offsite transmission grid) and Onsite Power Systems (AC and DC auxiliary system, both normal and safety related). In the chapter on the off-site system, attention is paid to the analysis of transmission capacity of the 400 kV lines, analysis of transient stability, multiple fault analyses, and probabilistic analyses of the grid and NPP power system reliability. In the chapter on the on-site system, attention is paid to the power balances of the electrical sources and switchboards set for various operational and accident modes, checks of loading and function of service and backup sources, short circuit current calculations, analyses of electrical protections, and analyses of the function and sizing of emergency sources (DG sets and UPS systems). (P.A.)

  2. The NPPR Trnava participation in the NPP V-2 modernisation and safety improvement project

    International Nuclear Information System (INIS)

    Michal, V.; Losonsky, B.; Magdolen, J.

    1999-01-01

    The presented contribution deals with form, present state and results of Nuclear Power Plants Research Inst.e participation in the NPP V-2 Jaslovske Bohunice Modernization and Safety Improvement Project.(author)

  3. Main results of substantiation of the ecological safety of the Novovoronezh NPP-2 design

    International Nuclear Information System (INIS)

    Kopytov, I.I.; Kocher'yan, V.M.; Leonov, S.V.; Chionov, V.G.; Ehrnestova, L.S.

    2005-01-01

    Paper presents the results of the efforts to determine both the actual (hydrochemical, hydrobiological, geobotanical, soil, radiological) and the predicted parameters of the region ecology derived when substantiating the ecological safety of the Novovoronezh NPP-2 design [ru

  4. NPP safety and personnel training. XII International conference. Abstracts. Volume 2

    International Nuclear Information System (INIS)

    2011-01-01

    The XII International conference NPP Safety and Personnel Training took place in Obninsk, October 4-7 2011. The problems of personnel training for nuclear industry are discussed. The innovation nuclear systems and fuel cycle are considered. The much attention has been given to NPP radiation safety and radioecology issues. The recent high-speed computation and simulation methods used in reactor technology are presented [ru

  5. Seismic safety programme at NPP Paks. Propositions for coordinated international activity in seismic safety of the WWER-440 V-213

    International Nuclear Information System (INIS)

    Katona, T.

    1995-01-01

    This paper presents the Paks NPP seismic safety program, highlighting the specifics of the WWER-440/213 type in operation, and the results of work obtained so far. It covers the following scope: establishment of the seismic safety program (original seismic design, current requirements, principles and structure of the seismic safety program); implementation of the seismic safety program (assessing the seismic hazard of the site, development of the new concept of seismic safety for the NPP, assessing the seismic resistance of the building and the technology); realization of the seismic safety of higher level (technical solutions, drawings, realization); ideas and propositions for coordinated international activity

  6. Joint probability safety assessment for NPP defense infrastructure against extreme external natural hazards

    International Nuclear Information System (INIS)

    Guilin, L.; Defu, L.; Huajun, L.; Fengqing, W.; Tao, Z.

    2012-01-01

    With the increasing tendency of natural hazards, the typhoon, hurricane and tropical Cyclone induced surge, wave, precipitation, flood and wind as extreme external loads menacing Nuclear Power Plants (NPP) in coastal and inland provinces of China. For all of planned, designed And constructed NPP the National Nuclear Safety Administration of China and IAEA recommended Probable Maximum Hurricane /Typhoon/(PMH/T), Probable Maximum Storm Surge (PMSS), Probable Maximum Flood (PMF), Design Basis Flood (DBF) as safety regulations for NPP defense infrastructures. This paper discusses the joint probability analysis of simultaneous occurrence typhoon induced extreme external hazards and compare with IAEA 2006-2009 recommended safety regulation design criteria for some NPP defense infrastructures along China coast. (authors)

  7. Assistance of Foreign Countries and International Organizations to Support Safety Improvements at Ignalina NPP

    International Nuclear Information System (INIS)

    Shevaldin, V.

    1997-01-01

    International cooperation and assistance for the improving safety of Ignalina NPP is described. Sweden was among the first countries which supported safety improvements at Ignalina NPP. The first project in the cooperation was BARSELINA, Probabilistic Safety Analysis of Ignalina NPP. The cooperation is still bringing significant support to the plant, including improvements in the fire protection, communications system, physical protection, and many other areas. Another one very important source of assistance was Nuclear Safety Account, administered by the EBRD. In 1993 experts of the plant, together with representatives of VATESI and SKI (Sweden) have worked out a short-term safety improvement program SIP-1, which was financed by the EBRD . Eighteen safety related projects were selected, expensive and reliable equipment was procured and installed

  8. Development of NPP Safety Requirements into Kenya's Grid Codes

    Energy Technology Data Exchange (ETDEWEB)

    Ndirangu, Nguni James; Koo, Chang Choong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand.

  9. Development of NPP Safety Requirements into Kenya's Grid Codes

    International Nuclear Information System (INIS)

    Ndirangu, Nguni James; Koo, Chang Choong

    2015-01-01

    As presently drafted, Kenya's grid codes do not contain any NPP requirements. Through case studies of selected grid codes, this paper will study frequency, voltage and fault ride through requirements for NPP connection and operation, and offer recommendation of how these requirements can be incorporated in the Kenya's grid codes. Voltage and frequency excursions in Kenya's grid are notably frequently outside the generic requirement and the values observed by the German and UK grid codes. Kenya's grid codes require continuous operation for ±10% of nominal voltage and 45.0 to 52Hz on the grid which poses safety issues for an NPP. Considering stringent NPP connection to grid and operational safety requirements, and the importance of the TSO to NPP safety, more elaborate requirements need to be documented in the Kenya's grid codes. UK and Germany have a history of meeting high standards of nuclear safety and it is therefore recommended that format like the one in Table 1 to 3 should be adopted. Kenya's Grid code considering NPP should have: • Strict rules for voltage variation, that is, -5% to +10% of the nominal voltage • Strict rules for frequency variation, that is, 48Hz to 52Hz of the nominal frequencyand

  10. Periodical safety review of units 1 and 2 of PAKS NPP. Examples from summary report

    International Nuclear Information System (INIS)

    Hammar, K.

    1998-01-01

    On the basis of American practice of qualification and relevant IAEA recommendations detailed guidelines of the qualification procedure were developed and executed on the Units 1 and 2 of the Paks NPP. Periodic safety supervision will be performed by evaluation of the following reports to be submitted by NPP: real technical conditions of the facility; existing practice and proposals for equipment qualification; evaluation of the existing safety reports estimating their validity up to the plant lifetime; ageing and ageing management; procedures of operation, maintenance, supervision; organisation and administration; safety impact of human factor, training, education, qualification of personnel

  11. Development of safety culture at Kozloduy NPP: The role of the regulatory body

    International Nuclear Information System (INIS)

    Miliovsky, Ventzislav

    2002-01-01

    An evolution is made upon external factor that accompany the development of safety culture at Kozloduy NPP (KNPP). The two basic stages and the major results achieved independently and with external assistance were considered in the process of development of safety culture. The areas in which the regulatory body insists on having active participation were defined. (author)

  12. Annual activity report of Ignalina NPP Safety Analysis Group for 1995 year

    International Nuclear Information System (INIS)

    Ushpuras, E.; Augutis, J.; Bubelis, E.

    1995-01-01

    The main results of Ignalina NPP Safety Analysis Group (ISAG) investigations for 1995 are presented. ISAG is concentrating its research activities into four areas: the neutrons dynamics modelling, simulation of transient processes during loss of coolant accident, the reactor cooling systems modelling and the probabilistic safety assessment of accident confinement system. 18 refs., 9 tabs., 110 figs

  13. Annual activity report of Ignalina NPP Safety Analysis Group for the year 1997

    International Nuclear Information System (INIS)

    Ushpuras, E.; Augutis, J.; Bubelis, E.; Kaliatka, A

    1998-01-01

    The main results of Ignalina NPP Safety Analysis Group (ISAG) investigations for the year 1997 are presented. ISAG is concentrating its research activities into four areas: the neutrons dynamics modelling, simulation of transient processes during loss of coolant accident, the reactor cooling systems modelling and the probabilistic safety assessment of accident confinement system

  14. UA criterion for NPP safety estimation SALP program

    International Nuclear Information System (INIS)

    Gorynina, L.V.; Tishchenko, V.A.

    1992-01-01

    Adopted by NRC program SALR is considered. The program is intended for acquisition and estimation of data on the activities of forms having licences for NPP operation and (or) construction. The criteria for estimation and the mechanism for determination of the rating of the firm activity quality are discussed

  15. Some research results by risk-inform approaches for NPP safety and operational efficiency

    International Nuclear Information System (INIS)

    Komarov, Yu.A.

    2013-01-01

    Article is the presentation of the same name monograph, which is planned to be issued. In the article the perspective problems of further development risk-oriented approach (ROA) for the grounding and realization of measures on increase of safety and operational efficiency of NPP are considered. Unlike the traditional approach for the ROA, mean due the definition of probabilistic and/or deterministic methods of risk parameters, as criterion functions essence and the measure of the estimation are defined by the solution of specific problem in nuclear field. The ROA application allows essentially expanding opportunities of the substantiations and realizations of measures on safety and operational efficiency increase of NPP

  16. Concept of space NPP radiation safety and its realization in the Kosmos-1900 satellite

    International Nuclear Information System (INIS)

    Gryaznov, G.M.; Nikolaev, V.S.; Serbin, V.I.; Tyugin, V.M.

    1989-01-01

    A standard NPP for a space vehicle, radioactivity composition and radiation safety systems are considered. Plausible accidents on board the space vehicle and requirements to system operation reliability are discussed. The main reactor characteristics situation on board the Kosmos-1900 satellite and completion of its flight are described. The experience in providing radiation safety of space NPP has shown that it is sufficient to use two independent systems: a drift system and a reactor dispersion system based on separation of its structure by active means

  17. Advances in safety countermeasures at the Tomari NPP of Hokkaido Electric Power on the basis of Fukushima Daiichi NPP accident. Fire protection and other advances

    International Nuclear Information System (INIS)

    Shibata, Taku; Dasai, Katsumi

    2014-01-01

    Fire protections for the nuclear power plants have been based on the fire laws and the conventional guide. After Fukushima Daiichi NPP accident, many safety countermeasures - also about Fire Protection - have been discussed in the Japanese authorities. This paper shows our present activities in the Tomari NPP about the fire protections from the view points of Fire Prevention, Fire Detection/Suppression Systems and Fire Protection, and other advances. (author)

  18. Annual activity report of Ignalina NPP Safety Analysis Group for 1996 year

    International Nuclear Information System (INIS)

    Ushpuras, E.; Augutis, J.; Bubelis, E.

    1997-03-01

    The main results of Ignalina NPP Safety Analysis Group (ISAG) investigations for 1996 are presented. ISAG is concentrating its research activities into four areas: the neutrons dynamics modelling, simulation of transient processes during loss of coolant accident, the reactor cooling systems modelling and the probabilistic safety assessment of accident confinement system. Ignalina Safety Analysis Report was prepared on the basis of these results. 37 refs., 9 tabs., 96 figs

  19. Use of the deterministic safety analyses in support to the NPP Krsko modification

    International Nuclear Information System (INIS)

    Feretic, D.; Cavlina, N.; Debrecin, N.; Grgic, D.; Bajs, T.; Spalj, S.

    2004-01-01

    The ultimate goal of the safety analysis is to verify that Nuclear Power Plant (NPP) meets safety and operational requirements. To this aim it is necessary to demonstrate that plant safety has not been deteriorated in the case of the modifications to the plant Systems, Structures and Components (SSC) or changes to the plant procedures. In addition, safety analyses are needed in the case of reassessment of an existing plant. The reasons for reassessment may be different, e.g. due to the changes in the methodology and assumptions used in the original design, if the original design basis or acceptance criteria may no longer be adequate, if the safety analysis tools used may have been superseded by more sophisticated methods or if the original design basis may no longer be met. The operation of the NPP Krsko has experienced numerous changes from the original design for the majority of the reasons that have been mentioned before. On the other side, the application of the large best-estimate thermalhydraulic codes has evolved to the wide spread support in the operation of the NPP: compliance with the regulatory goals, support to the PSA studies, analysis of the operational transients, plant modifications studies, equipment qualification, training of the operators, preparation of the operating procedures, etc. This trend has been followed at the Faculty of Electrical Engineering Zagreb (FER) and applied to the on-going needs due to the modifications and changes at NPP Krsko. In this paper, an overview of the deterministic safety analyses performed at FER in the support to the NPP Krsko modifications and changes is presented.(author)

  20. Generalization of Nuclear Safety and Course of Accident Events Research in the Ignalina NPP

    International Nuclear Information System (INIS)

    Kaliatka, A.; Uspuras, E.

    2001-01-01

    The safety analysis shown that after implementation of SAR recommendations Ignalina NPP is adequately protected against accidents which required fast initiation of automatic protections. In case of accidents with long-term loss of core cooling additional operator actions are required. Accident management in case long-term core cooling are analyzed in this paper. (author)

  1. Annual activity report of Ignalina NPP Safety Analysis Group for 1994 year

    International Nuclear Information System (INIS)

    Ushpuras, E.; Kaliatka, A.; Chesna, B.; Dundulis, G.

    1995-01-01

    The main results of Ignalina NPP Safety Analysis Group (ISAG) investigations for 1994 are presented. ISAG is concentrated its research activities into 3 areas: the neutrons dynamics modeling, simulation of transient processes during loss of coolant accident and calculation of reactor building structure's streses and other mechanical properties in the case of accident. 6 refs., 13 tabs., 69 figs

  2. Experience with Periodic Safety Review (PSR) at Kozloduy NPP after 20 years of operation

    International Nuclear Information System (INIS)

    Popov, V.

    2011-01-01

    Conclusion: Measures in the long-term Programs for improving safety and radiation protection of unit 5&6, based on PSR outcome and addressed on the units’ operation licence renewal are under way. • NPP Kozloduy intends to fulfils requirements on modern nuclear power plants which occasionally exceeds limits of the effective national nuclear legislation

  3. Maintenance optimization of the Ignalina NPP safety systems

    International Nuclear Information System (INIS)

    Zuokas, D.; Augutis, J.

    2000-01-01

    This paper represents some theses and mathematical approach to optimization of technical specification problems of the diesel generators system (DGS) at the Ignalina NPP (INPP) in Lithuania. Analysis unites three chapters, each attributed to the solving of related problems: dependency of the Core Damage Frequency (CDF) on DGS test interval (Chapter 2), analysis and comparison of two different maintenance strategies for DGS (Chapter 3), DGS test interval's optimisation (averaged loss function minimisation) taking into consideration maintenance and DGS failure's consequence costs (Chapter 4). The mathematical model was developed which lets to estimate DGS failure probability, the averaged loss function value and study the influence of different parameters into final results. (author)

  4. An approach for Periodic Safety Review (PSR) of units 5 and 6 of Kozloduy NPP

    International Nuclear Information System (INIS)

    Kichev, Emil

    2014-01-01

    Periodic Safety Reviews (PSR) is complementary to the routine and special safety reviews and does not replaced them. It determines the level of compliance with current safety requirements and covers all safety aspects of NPP operation and all equipment (SSCs) on site. Three levels of importance are defined for the documents in PSR (towards higher conservatism): Bulgarian documents; IAEA documents; Documents of manufactures and suppliers of the equipment (SSCs). The scope and activities consists of four stages and includes 14 safety factors (SF). Cooperation with external companies and organizations in areas with proved experience is used.

  5. Evaluation of the safety margins during shutdown for NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Sadek, S.; Bajs, T.

    2004-01-01

    In the paper the results of RELAP5/mod3.3 calculations of critical parameters during shutdown for NPP Krsko are presented. Conservative evaluations have been performed at NPP Krsko to determine the minimum configuration of systems required for the safe shutdown operation. Critical parameters in these evaluations are defined as the time to start of the boiling and the time of the core dry-out. In order to have better insight into the available margins, the best estimate code RELAP5/mod3.3 has been used to calculate the same parameters. The analyzed transient is the loss of the Residual Heat Removal (RHR) system, which is used to remove decay heat during shutdown conditions. Several configurations that include open and closed Reactor Coolant System (RCS) were considered in the evaluation. The RELAP5/mod3.3 analysis of the loss of the RHR system has been performed for the following cases: 1) RCS closed and water solid, 2) RCS closed and partially drained, 3) Pressurizer manway open, Steam Generator (SG) U tubes partially drained, 4) Pressurizer and SG manways open, SG U tubes completely drained, 5) Pressurizer manway open, SGs drained, SG nozzle dams installed and 6) SG nozzle dams installed, pressurizer manway open, 1 inch break at RHR pump discharge in the loop with pressurizer. Both RHR trains were assumed in operation prior to start of the transient. The maximum average steady state temperature for all analyzed cases was limited to 333 K. (author)

  6. Screening of external hazards for NPP with bank type reactor. Modeling of safety related systems and equipment for RBMK. Probabilistic assessment of NPP safety on aircraft impact. Progress report

    International Nuclear Information System (INIS)

    Kostarev, V.

    1999-01-01

    This progress report was produced within the frame of IAEA research project on screening the hazards for NPP with bank type reactor. It covers the following tasks; development of the model for the primary loop system of RBMK; developing the models for safety related equipment of RBMK; developing of models for safety related models of EGP-6 type reactor (Bilibinskaya Nuclear Co-generated heat and Power Plant); and probabilistic assessment of NPP safety on aircraft impact

  7. Safety critical FPGA-based NPP instrumentation and control systems: assessment, development and implementation

    Energy Technology Data Exchange (ETDEWEB)

    Bakhmach, E. S.; Siora, A. A.; Tokarev, V. I. [Research and Production Corporation Radiy, 29 Geroev Stalingrada Str., Kirovograd 25006 (Ukraine); Kharchenko, V. S.; Sklyar, V. V.; Andrashov, A. A., E-mail: marketing@radiy.co [Center for Safety Infrastructure-Oriented Research and Analysis, 37 Astronomicheskaya Str., Kharkiv 61085 (Ukraine)

    2010-10-15

    The stages of development, production, verification, licensing and implementation methods and technologies of safety critical instrumentation and control systems for nuclear power plants (NPP) based on FPGA (Field Programmable Gates Arrays) technologies are described. A life cycle model and multi-version technologies of dependability and safety assurance of FPGA-based instrumentation and control systems are discussed. An analysis of NPP instrumentation and control systems construction principles developed by Research and Production Corporation Radiy using FPGA-technologies and results of these systems implementation and operation at Ukrainian and Bulgarian NPP are presented. The RADIY{sup TM} platform has been designed and developed by Research and Production Corporation Radiy, Ukraine. The main peculiarity of the RADIY{sup TM} platform is the use of FPGA as programmable components for logic control operation. The FPGA-based RADIY{sup TM} platform used for NPP instrumentation and control systems development ensures sca lability of system functions types, volume and peculiarities (by changing quantity and quality of sensors, actuators, input/output signals and control algorithms); sca lability of dependability (safety integrity) (by changing a number of redundant channel, tiers, diagnostic and reconfiguration procedures); sca lability of diversity (by changing types, depth and method of diversity selection). (Author)

  8. Safety critical FPGA-based NPP instrumentation and control systems: assessment, development and implementation

    International Nuclear Information System (INIS)

    Bakhmach, E. S.; Siora, A. A.; Tokarev, V. I.; Kharchenko, V. S.; Sklyar, V. V.; Andrashov, A. A.

    2010-10-01

    The stages of development, production, verification, licensing and implementation methods and technologies of safety critical instrumentation and control systems for nuclear power plants (NPP) based on FPGA (Field Programmable Gates Arrays) technologies are described. A life cycle model and multi-version technologies of dependability and safety assurance of FPGA-based instrumentation and control systems are discussed. An analysis of NPP instrumentation and control systems construction principles developed by Research and Production Corporation Radiy using FPGA-technologies and results of these systems implementation and operation at Ukrainian and Bulgarian NPP are presented. The RADIY TM platform has been designed and developed by Research and Production Corporation Radiy, Ukraine. The main peculiarity of the RADIY TM platform is the use of FPGA as programmable components for logic control operation. The FPGA-based RADIY TM platform used for NPP instrumentation and control systems development ensures sca lability of system functions types, volume and peculiarities (by changing quantity and quality of sensors, actuators, input/output signals and control algorithms); sca lability of dependability (safety integrity) (by changing a number of redundant channel, tiers, diagnostic and reconfiguration procedures); sca lability of diversity (by changing types, depth and method of diversity selection). (Author)

  9. Enhanced safety features of CHASHMA NPP UNIT-2 to encounter selected severe accidents, various challenges involved to prove the adequacy of severe accidents prevention/mitigation measures and to write management guidelines with one possible solution to these challenges

    International Nuclear Information System (INIS)

    Iqbal, Z.; Minhaj, A.

    2007-01-01

    This paper describes enhanced safety features of Chashma Nuclear Power Plant Unit-2 (C-2), a 325 MWe PWR to encounter selected severe accidents and discusses various challenges involved to prove the adequacy of severe accidents encountering measures and to write severe accident management guidelines (SAMGs) in compliance with the recently introduced national regulations based on the new IAEA nuclear safety standards. C-2 is being built by China National Nuclear Corporation (CNNC) for Pakistan Atomic Energy Commission (PAEC). Its twin, Unit-1 (C-1) also a 325 MWe PWR, was commissioned in 2000. Nuclear power safety with reference to severe accidents should be treated as a global issue and therefore the developed countries should include the people of developing countries in nuclear power industry's various severe accidents based research and development programs. The implementation of this idea may also deliver few other useful and mutually beneficial byproducts. (author)

  10. Development of Safety Culture Assessment Strategy for Korean NPP

    International Nuclear Information System (INIS)

    Park, Jung Hwan; Kim, Jong Hyun

    2014-01-01

    This paper aims at developing the requirements for a method to evaluate the operational safety culture, evaluating currently available methods based on the requirements, and suggesting a method to evaluate and improve the operational safety culture for Korean nuclear power plants. This paper reviews the widely-used methods to assess safety culture for NPPs and their basis. Then, this paper develops the requirements for the method to evaluate operational safety culture for Korean NPPs. Based on these requirements, Korean Safety Culture Indicators (KSCI) and evaluation measures are also suggested. Finally this paper proposes the guidelines to develop improvements to safety culture from the evaluation results

  11. Development of Safety Culture Assessment Strategy for Korean NPP

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jung Hwan; Kim, Jong Hyun [KEPCO, Ulsan (Korea, Republic of)

    2014-08-15

    This paper aims at developing the requirements for a method to evaluate the operational safety culture, evaluating currently available methods based on the requirements, and suggesting a method to evaluate and improve the operational safety culture for Korean nuclear power plants. This paper reviews the widely-used methods to assess safety culture for NPPs and their basis. Then, this paper develops the requirements for the method to evaluate operational safety culture for Korean NPPs. Based on these requirements, Korean Safety Culture Indicators (KSCI) and evaluation measures are also suggested. Finally this paper proposes the guidelines to develop improvements to safety culture from the evaluation results.

  12. Equipment maintenance as a tool for raising of safety and efficiency of the NPP'Kozloduy' operation

    International Nuclear Information System (INIS)

    Kichev, E.

    1999-01-01

    The role of the maintenance in the NPP management is discussed. The main links in the management are: NPP executives and their position; maintenance; quantitative characteristics of the maintenance. The described management process gives the opportunity the maintenance optimization on different levels: component, system, unit, plant. The results are valid for other areas with high safety requirements

  13. Role of supervising authorities in NPP operation safety ensuring

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2001-01-01

    The main working spheres and results gained during 40 years of activity of the Great Britain Nuclear Installation Inspectorate (NII) are considered. The new approach to safety analysis developed with NII participation is described in details. The important role of the safety analysis realization, utilization of modern methods for risk estimation and safety culture principles introduction at NPPs is shown [ru

  14. Choice and complexation of techniques and tools for assessment of NPP I and C systems safety

    International Nuclear Information System (INIS)

    Illiashenko, Oleg; Babeshko, Eugene

    2011-01-01

    There are a lot of techniques to analyze and assess reliability and safety of NPP Instrumentation and Control (I and C) systems (e.g. FMEA - Failure Modes and Effects Analysis and its modifications, FTA - Fault Tree Analysis, HAZOP - Hazard and Operability Analysis, RBD - Reliability Block Diagram, Markov Models, etc.) and quantity of tools based on these techniques is constantly increasing. Known ways of safety assessment, as well as problems of their choice and complexation are analyzed. Objective of the paper is the development of general 'technique of techniques choosing' and tool for support of such technique. The following criteria are used for analysis and comparison and their features are described: compliance to normative documents; experience of application in industry; methods used for assessment of system NPP I and C safety; tool architecture/framework; reporting; vendor support, etc. Comparative analysis results of existing T and T - Tools and Techniques for safety analysis are presented in matrix form ('Tools-Criterion') with example. Features of complexation of different safety assessment techniques (FMECA, FTA, RBD, Markov Models) are described. The proposed technique is implemented as special tool for decision-making. The proposed technique was used for development of RPC Radiy company standard CS 66. This guide contains requirements and procedures of FMECA analysis of developed and produced NPP I and C systems based on RADIY platform. (author)

  15. Research on Integration of NPP Operational Safety Management Performance Systems

    International Nuclear Information System (INIS)

    Chi, Miao; Shi, Liping

    2014-01-01

    The operational safety management of Nuclear Power Plants demands systematic planning and integrated control. NPPs are following the well-developed safety indicator systems proposed by IAEA Operational Safety Performance Indicator Programme, NRC Reactor Oversight Process or the other institutions. Integration of the systems is proposed to benefiting from the advantages of both systems and avoiding improper application into the real world. The authors analyzed the possibility and necessity for system integration, and propose an indicator system integrating method

  16. Human Resources Readiness as TSO for Deterministic Safety Analysis on the First NPP in Indonesia

    International Nuclear Information System (INIS)

    Sony Tjahyani, D. T.

    2010-01-01

    In government regulation no. 43 year 2006 it is mentioned that preliminary safety analysis report and final safety analysis report are one of requirements which should be applied in construction and operation licensing for commercial power reactor (NPPs). The purpose of safety analysis report is to confirm the adequacy and efficiency of provisions within the defence in depth of nuclear reactor. Deterministic analysis is used on the safety analysis report. One of the TSO task is to evaluate this report based on request of operator or regulatory body. This paper discusses about human resources readiness as TSO for deterministic safety analysis on the first NPP in Indonesia. The assessment is done by comparing the analysis step on SS-23 and SS-30 with human resources status of BATAN currently. The assessment results showed that human resources for deterministic safety analysis are ready as TSO especially to review preliminary safety analysis report and to revise final safety analysis report in licensing on the first NPP in Indonesia. Otherwise, to prepare the safety analysis report is still needed many competency human resources. (author)

  17. Seismic analysis of the safety related piping and PCLS of the WWER-440 NPP

    International Nuclear Information System (INIS)

    Berkovski, A.M.; Kostarev, V.V.; Schukin, A.J.; Boiadjiev, Z.; Kostov, M.

    2001-01-01

    This paper presents the results of seismic analysis of Safety Related Piping Systems of the typical WWER-440 NPP. The methodology of this analysis is based on WANO Terms of Reference and ASME BPVC. The different possibilities for seismic upgrading of Primary Coolant Loop System (PCLS) were considered. The first one is increasing of hydraulic snubber units and the second way is installation of limited number of High Viscous Dampers (HVD). (author)

  18. Two important safety-related verification tests in the design of Qinshan NPP 600 MWe reactor

    International Nuclear Information System (INIS)

    Li Pengzhou; Li Tianyong; Yu Danping; Sun Lei

    2005-01-01

    This paper summarizes two most important verification tests performed in the design of reactor of Qinshan NPP Phase II: seismic qualification test of control rod drive line (CRDL), flow-induced vibration test of reactor internals both in 1:5 scaled model and on-site measurement during heat function testing (HFT). Both qualification tests proved that the structural design of the reactor has large safety margin. (authors)

  19. Operation and safety decision-making support expert system in NPP

    International Nuclear Information System (INIS)

    Wei Yanhui; Su Desong; Chen Weihua; Zhang Jianbo

    2014-01-01

    The article first reviewed three operation support systems currently used in NPP: real-time information surveillance system, important equipment surveillance system and plant process control and monitoring system, then presents the structure and function of three expert support sub-systems (intelligent alarm monitoring system, computer-based operating procedure support system, safety information expert decision support system). Finally the article discussed the meaning of a kind of operation decision making support system. (authors)

  20. Establishing the Safety Infrastructure for NPP in Mongolia

    Energy Technology Data Exchange (ETDEWEB)

    Enkhbat, Norov [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of); Lee, Y. E. [Korean Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    The survey results and analyses were used to establish the basis for developing nuclear safety infrastructure in Mongolia. Power Energy effective reform or introduction of nuclear reactors should be implemented in coming future to avoid this critical situation faces us. The most participant suggested that Mongolia may cooperate in the field of nuclear safety and infrastructure development with the Republic of Korea. Nuclear Energy Agency of the government of Mongolia has organized Nuclear Power Infrastructure Development (NUPID) training in cooperating with Seoul National University and other organizations in 2008, 2010 and 2012. There is a need to improve the nuclear energy law of Mongolia. Total energy supply of Mongolia is 5124.08 MWt as of 2012. 92.4% of total energy supply produces with coal, 0.56% with liquid fuel, and 0.01% with renewable energy sources, remained 6% imports from Russia. Mongolia operates seven Thermal Centralized Systems (TCS) with total capacity of 802 MWt, which provides dual; electricity and thermal power. Energy demand in Mongolia is expected to increase, due to the dramatically expanding mining industry. It is absolutely impossible to supply such rapid growth having operated old technology and inefficient production which exists currently in Mongolia. Therefor Mongolian government is interested in utilizing nuclear energy and approved Nuclear Energy Law in 2009. National Development Strategy (2008-2021) stated as that the peaceful exploitation of the nuclear energy will be an important factor for the sustainable development of Magnolia. Action Plan of the Government for 2008-2012 stated as that ...conduct a comprehensive research for use of nuclear energy, develop technical and economic feasibility study and improve radiation control and safety. International community has developed appropriate approaches in the form of IAEA safety standards, which has a positive experience of regulation and safety. These approaches contribute to the

  1. Establishing the Safety Infrastructure for NPP in Mongolia

    International Nuclear Information System (INIS)

    Enkhbat, Norov; Lee, Y. E.

    2013-01-01

    The survey results and analyses were used to establish the basis for developing nuclear safety infrastructure in Mongolia. Power Energy effective reform or introduction of nuclear reactors should be implemented in coming future to avoid this critical situation faces us. The most participant suggested that Mongolia may cooperate in the field of nuclear safety and infrastructure development with the Republic of Korea. Nuclear Energy Agency of the government of Mongolia has organized Nuclear Power Infrastructure Development (NUPID) training in cooperating with Seoul National University and other organizations in 2008, 2010 and 2012. There is a need to improve the nuclear energy law of Mongolia. Total energy supply of Mongolia is 5124.08 MWt as of 2012. 92.4% of total energy supply produces with coal, 0.56% with liquid fuel, and 0.01% with renewable energy sources, remained 6% imports from Russia. Mongolia operates seven Thermal Centralized Systems (TCS) with total capacity of 802 MWt, which provides dual; electricity and thermal power. Energy demand in Mongolia is expected to increase, due to the dramatically expanding mining industry. It is absolutely impossible to supply such rapid growth having operated old technology and inefficient production which exists currently in Mongolia. Therefor Mongolian government is interested in utilizing nuclear energy and approved Nuclear Energy Law in 2009. National Development Strategy (2008-2021) stated as that the peaceful exploitation of the nuclear energy will be an important factor for the sustainable development of Magnolia. Action Plan of the Government for 2008-2012 stated as that ...conduct a comprehensive research for use of nuclear energy, develop technical and economic feasibility study and improve radiation control and safety. International community has developed appropriate approaches in the form of IAEA safety standards, which has a positive experience of regulation and safety. These approaches contribute to the

  2. Performance indicators at Embalse NPP: PSA and safety system indicators based on PSA models

    International Nuclear Information System (INIS)

    Fornero, D.A.

    2001-01-01

    Several indicators have been implemented at Embalse NPP. The objective was selecting some representative parameters to evaluate the performance of both the plant and the personnel activities, important for safety. A first set of indicators was defined in accordance with plant technical staff criteria. A complementary set of them was addressed later based on WANO guidance. This report presents the set of indicators used at Embalse NPP, centering the description to related to safety systems performance indicators (SSPI). Some considerations are done about the calculation methods, the need for aligning and updating their values following Embalse Probabilistic Safety Assessment (PSA) development, and some pros and cons of using the PSA model for getting systems indicators. Owing to the fact that PSA ownership by utilities is also a subject of the meeting, some characteristics of the organization of the PSA Project are described at the beginning of the report. At Embalse NPP a Level 1 PSA has been developed under the responsibility of its own plant and with an important contribution from the IAEA. PSA was developed at the site, conducting this to a study strongly interactive with the station staff. (author)

  3. Indicators to monitor NPP safety performance. Working material

    International Nuclear Information System (INIS)

    1997-01-01

    Numerical indicators to monitor safety status and overall safety performance of nuclear power plants (NPPs) are used by operators and some regulators worldwide. During the last few years, the IAEA, through Technical Committee Meetings and Consultants' Meetings has worked on this area. This report presents a framework for nuclear power plant safety performance indicators that was developed during two consultant meetings held at the IAEA headquarters in Vienna in December 1995 and November 1996. Annex 2 on risk based indicators was prepared during a consultants' meeting held in Vienna in July 1996. An additional outcome from these activities, was the recommendation that the IAEA conduce pilot exercises at several nuclear power plants that might be interested to participate, in order to test the validity of the concept and its usefulness. 6 figs

  4. Design on human supervisory control of safety management for advance NPP

    International Nuclear Information System (INIS)

    Nugroho, D.H.; Soentono, S.; Taryo, T.; Wahyon, P.I.

    2006-01-01

    Full text: Full text: Characteristics of an advance NPP related with economic and safety enhancement was represented on capabilities in intelligent control and diagnostic to provide nearly autonomous operation with anticipatory maintenance. An autonomous control system should enable automatic operation while adapting to component faults and system variable upsets. It needs therefore to have many intelligent capabilities, such as modeling, analysis, self-validation, diagnosis and decision. This paper describes a human supervisory control design for nuclear reactor fault management which collaborates between human and autonomous control. The beneficial of collaboration is provided based on belief of information value evaluated from Dempster's rule of evidence. The belief of the collaboration is better compared with single usage. The collaboration was conducted in which agent will autonomously and periodically be conducting surveillance by checking the component abnormalities in the plant if anomalies occur. The anomalies were determined using fault detection module. Thereby the system will be able to conduct preventive maintenance. In the situation of accident happened, hence the system will diagnose to determine the location of component failure autonomously. A human supervisor will then determine the action of decison making based on the prediction result. The decision making will be conducted based on the 4th Sheridan's autonomous level in which the autonomous control will execute the duty autonomously when the plant is in normal condition, or in the predictable accident range. But if the unpredictable accident occurs in the plant, the supervisor will then take over the role to control the plant, and the machine will do what commanded by the supervisor. Sistematically, the system explained before was represented by Traveling Salesman Problem-based surveillance, modified ART-2 artificial neural networks-based fault detection and Bayesian Networks-based fault

  5. Expert evaluation in NPP safety important systems licensing process

    International Nuclear Information System (INIS)

    Mikhail, A Yastrebenetsky; Vasilchenko, V.N.

    2001-01-01

    Expert evaluation of nuclear power plant safety important systems modernization is an integral part of these systems licensing process. The paper contains some aspects of this evaluation which are based on Ukrainian experience of VVER-1000 and VVER-440 modernization. (authors)

  6. Expert evaluation in NPP safety important systems licensing process

    Energy Technology Data Exchange (ETDEWEB)

    Mikhail, A Yastrebenetsky; Vasilchenko, V.N. [Ukrainian State Scientific Technical Center of Nuclear and Radiation Safety (Ukraine)

    2001-07-01

    Expert evaluation of nuclear power plant safety important systems modernization is an integral part of these systems licensing process. The paper contains some aspects of this evaluation which are based on Ukrainian experience of VVER-1000 and VVER-440 modernization. (authors)

  7. The development of NPP operational safety training courses

    International Nuclear Information System (INIS)

    Lee, Chang Kun; Lee, Duk Sun; Lee, Byung Sun; Lee, Won Koo; Juhn, Heng Run; Moon, Byung Soo; Cho, Min Sik; Lee, Han Young; Moon, Hak Won; Seo, Yeon Ho

    1987-12-01

    The objective of the project is to develop a training course text for the betterment of reactor operation and assurance of its safety in general by providing training materials of the advanced compact nuclear simulator which will become operation in September 1988. Main scope and contents of the project are as follows: - compilation of basic data related to simulator operation and maintenance as well as the comparative analysis with respect to simulator materials in foreign countries - method of training by simulator - review the training status by simulator in foreign countries - development of training course in the field of reactor safety It is expected that the results will be reflected to the actual training and retraining of the reactor operating crew so as to improve and update their capabilities in training fashion. (Author)

  8. Criticality safety analysis of spent fuel storage for NPP Mochovce using MCNP5

    International Nuclear Information System (INIS)

    Farkas, G.; Hascik, J.; Lueley, J.; Vrban, B.; Petriska, M.; Slugen, V.; Urban, P.

    2011-01-01

    The paper presents results of nuclear criticality safety analysis of spent fuel storage for the first and second unit of NPP Mochovce. The spent fuel storage pool (compact and reserve grid) was modeled using the Monte Carlo code MCNP5. Conservative approach was applied and calculation of k eff values was performed for normal and various postulated emergency conditions in order to evaluate the final maximal k eff values. The requirement of current safety regulations to ensure 5% subcriticality was met except one especially conservative case. (Authors)

  9. Improved safety features in the design of Alto Lazio NPP

    International Nuclear Information System (INIS)

    Bava, G.; Cianciolo, T.; Del Nero, G.

    1988-01-01

    The ALTO LAZIO Nuclear Power Plant, two 1000Mwe units, is a BWR 6/MARK III located about 100 km north of Rome, on the Tyrrhenian Sea Coasts. The construction of the plant started in 1978, but it has recently been stopped by a Government decision following a national referendum, when the units were about 70% completed. This paper is mainly intended to illustrate the major safety features which have been implemented as result of specific requirements issued by the safety authority (ENEA DISP) during the construction permit stage or the subsequent licensing process. One of the tools used to identify the need for design modifications has been a comprehensive reliability analysis of safety system: in the paper the methods used and the major results obtained by this study are briefly presented. Also, the approach used in the investigation of severe accidents and major applications in the area of plant design and emergency procedures are briefly discussed; furthermore the trend toward a simpler mitigation concept is described

  10. Seismic analysis for safety related structures of 900MWe PWR NPP

    International Nuclear Information System (INIS)

    Liu Wei

    2002-01-01

    Nuclear Power Plant aseismic design becomes more and more important in China due to the fact that China is a country where earthquakes occur frequently and most of plants arc unavoidably located in seismic regions. Therefore, Chinese nuclear safety authority and organizations have worked out a series of regulations and codes related to NPP anti-seismic design taking account of local conditions. The author presents here an example of structural anti-seismic design of 90GM We PWR NPP which is comprised of: ground motion input, including the principles for ground motion determination and time history generation; soil and upper-structure modelling, presenting modeling procedures and typical models of safety related buildings such as Reactor Building, Nuclear Auxiliary Building and Fuel Building; soil-structure interaction analysis; and in-structure response analysis and floor response spectrum generation. With this example, the author intends to give an overview of Chinese practice in NPP structure anti-seismic design such as the main procedures to be followed and the codes and regulations to be respected. (author)

  11. Assessment of safety measures and plant risks during shut-down periods at NPP Biblis

    International Nuclear Information System (INIS)

    Pamme, H.; Roess, P.H.

    1996-01-01

    Results of French probabilistic PWR-studies indicated a high contribution of shut down states to the overall plant risk. Mainly inspired by these results qualitative and also probabilistic analyses were started in Germany since 1991. As the Biblis-B-NPP was already reference plant of the German Risk Study first studies of shut-down-states were again focuses on Biblis-B. These studies were mainly performed by the German Association for Reactor Safety (GRS) in close cooperation with the utility RWE Energie AG. This paper briefly reviews the chosen approach to model and assess shut-down-states at Biblis-NPP. An in-depth-presentation is focuses on the quantification of risk in mid-loop-operation (MLO) which was performed by the authors with intensive support of plant personnel

  12. Reliability Analysis on NPP's Safety-Related Control Module with Field Data

    International Nuclear Information System (INIS)

    Lee, Sang Yong; Jung, Jae Hyun; Kim, Seong Hun

    2006-01-01

    The automatic control systems used in nuclear power plant (NPP) consists of numerous control modules that can be considered to be a network of components various complex ways. The control modules require relatively high reliability than industrial electronic products. Reliability prediction provides the rational basis of system designs and also provides the safety significance of system operations. The aim of this paper is to minimize the deficiencies of the traditional reliability prediction method calculation using the available field return data. This way is possible to do more realistic reliability assessment. SAMCHANG Enterprise Company (SEC) has established database containing high quality data at the module and component level from module maintenance in NPP. On the basis of these, this paper compares results that add failure record (field data) to Telcordia-SR-332 reliability prediction model with MIL-HDBK-217F prediction results

  13. Substantiation of the safety in the technical project of Belene NPP

    International Nuclear Information System (INIS)

    Boyadzhiev, A.

    1990-01-01

    The chapter contains an evaluation of the safety of Belene NPP project, based on an experts study of the corresponding volume of the Technical Project documentation of the main contractor and also on other related documents. The authors state that most of the remarks are constitutive, part of them requiring additional information or research. The general explicit conclusion is that the materials on the safety substantiation provided in the project are insufficient for making final statements on the safety of the NPP and there is a need for a detailed analysis and expertise. There are 12 topical conclusion paragraphs and each of them comprises a number of remarks. Among the remarks there are some related to the reactivity coefficient values in certain modes of operation, the problem of the mechanical safety and control system efficiency, the unacceptable operation at nominal power in case of stringent safety rules enforcement, the insufficiency of the PSA, the automatic control systems and the software codes not standing up to the contemporary requirements. (R.Ts.)

  14. Conceptual Design of Hybrid Safety Features for NPP by Utilizing Solar Updraft Tower

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of); Choi, Young Jae; Kim, Yong Jin [KAIST, Daejeon (Korea, Republic of); Park, Hyo Chan; Park, Youn Won [BEES, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, hybrid safety features for NPP with solar updraft tower (SUT) is conceptually suggested to cope with loss of ultimate heat sink accident. The hybrid safety features utilizing SUT target NPPs in seashore of Arabian Gulf. Usually NPPs are constructed near seashore to utilize sea water as an ultimate heat sink. Residual heat or decay heat of nuclear reactor will diffuse into the ocean through the condenser. NPPs in Middle East are expected to be placed in seashore of Arabian Gulf. The NPP site of Barakah is an actual example. For NPPs in seashore of Arabian Gulf, an additional safety concern should be considered. Arabian Gulf is the largest oil transporting route in the world. The oil spill risk in Arabian Gulf will be the largest simultaneously. Unfortunately, not like other oceans, Arabian Gulf is a kind of closed ocean which does not have strong ocean currents connected to out of the gulf. If once oil spill is occurred, its influence can be propagated more than our expectation. The spilled oil also can affect to NPPs in seashore by covering surfaces of condenser. It will directly cause loss of ultimate heat sink. The hybrid safety features of SUT system are expected to aid normal operation of safety system and mitigate consequence of severe accident. Detail analysis and technology development is ongoing now.

  15. Verification and validation issues for digitally-based NPP safety systems

    International Nuclear Information System (INIS)

    Ets, A.R.

    1993-01-01

    The trend toward standardization, integration and reduced costs has led to increasing use of digital systems in reactor protection systems. While digital systems provide maintenance and performance advantages, their use also introduces new safety issues, in particular with regard to software. Current practice relies on verification and validation (V and V) to ensure the quality of safety software. However, effective V and V must be done in conjunction with a structured software development process and must consider the context of the safety system application. This paper present some of the issues and concerns that impact on the V and V process. These include documentation of systems requirements, common mode failures, hazards analysis and independence. These issues and concerns arose during evaluations of NPP safety systems for advanced reactor designs and digital I and C retrofits for existing nuclear plants in the United States. The pragmatic lessons from actual systems reviews can provide a basis for further refinement and development of guidelines for applying V and V to NPP safety systems. (author). 14 refs

  16. Conceptual Design of Hybrid Safety Features for NPP by Utilizing Solar Updraft Tower

    International Nuclear Information System (INIS)

    Song, Sub Lee; Choi, Young Jae; Kim, Yong Jin; Park, Hyo Chan; Park, Youn Won

    2016-01-01

    In this study, hybrid safety features for NPP with solar updraft tower (SUT) is conceptually suggested to cope with loss of ultimate heat sink accident. The hybrid safety features utilizing SUT target NPPs in seashore of Arabian Gulf. Usually NPPs are constructed near seashore to utilize sea water as an ultimate heat sink. Residual heat or decay heat of nuclear reactor will diffuse into the ocean through the condenser. NPPs in Middle East are expected to be placed in seashore of Arabian Gulf. The NPP site of Barakah is an actual example. For NPPs in seashore of Arabian Gulf, an additional safety concern should be considered. Arabian Gulf is the largest oil transporting route in the world. The oil spill risk in Arabian Gulf will be the largest simultaneously. Unfortunately, not like other oceans, Arabian Gulf is a kind of closed ocean which does not have strong ocean currents connected to out of the gulf. If once oil spill is occurred, its influence can be propagated more than our expectation. The spilled oil also can affect to NPPs in seashore by covering surfaces of condenser. It will directly cause loss of ultimate heat sink. The hybrid safety features of SUT system are expected to aid normal operation of safety system and mitigate consequence of severe accident. Detail analysis and technology development is ongoing now

  17. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu. N.; Lisitsa, F. D.; Smirnov, V. G.

    2000-01-01

    The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper three-quarter erection of steam-gas toppings to the nuclear power units three-quarter is considered in the paper. Application of the steam-gas toppings permits through reducing power of aging reactors to extend lifetime of nuclear power unit, enhance its safety and at the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project. for Russian boiling VK-50 reactor now in operation. Heat flow sheets of the power plants, their parameters and economic problems are discussed. (author)

  18. Use of a Computerized Tool (ORAM) to Help Manage Outage Safety and Risk at NPP Krsko

    International Nuclear Information System (INIS)

    Spiler, J.; Basic, I.; Vrbanic, I.; Fifnja, I.; Kastelan, M.; Dagan, W. J.; Shanley, L. B.; Naum, T. J.

    1998-01-01

    Outage Risk Assessment and Management (ORAM) is a computerized methodology developed by the U.S. Electric Power Research Institute (EPRI) to help Nuclear Power Plant personnel manage the risk and safety associated with refueling and forced plant outages. Today, over 60 plants including NPP Krsko are using ORAM during the preparation and performance of plant outages. In fact, many plants are attributing much of the reductions in the duration of refueling outages to the use of ORAM. The success of the ORAM methodology is the capability to provide plant and management personnel with understandable results from both deterministic evaluations of plant safety and quantitative risk assessments. The Nuklearna Elektrarna Krsko (NEK) use of ORAM involves both of these approaches. The deterministic portion of ORAM is used to model the NPP Krsko Shutdown Technical Specifications and administrative considerations. The probabilistic portion of ORAM uses industry and NEK specific initiating events and other risk elements pertaining to shutdown to derive a quantitative risk assessment for various end states, including core damage and RCS boiling. This paper expands on the value of each approach and demonstrates the benefits of combining these elements in the decision-making process. Another key advantage of ORAM is the ability to apply the methodology to specific outages. Since no outage is identical, this provides tremendous benefits to plant personnel for managing the safety and risk of a particular outage. ORAM does this ba organizing all of the various plant configurations and equipment unavailability windows into numerous plant states. Furthermore, ORAM evaluations can be a utomated b y interfacing with outage scheduling software programs such as Primavera. For each plant state, the deterministic and the probabilistic logic evaluations are applied. This paper will demonstrate the ORAM evaluation for an actual NPP Krsko outage. (author)

  19. Research and practice on NPP safety DCS application software V and V defect classification system

    International Nuclear Information System (INIS)

    Zhang Dongwei; Li Yunjian; Li Xiangjian

    2012-01-01

    One of the most significant aims of Verification and Validation (V and V) is to find software errors and risks, especially for a DCS application software designed for nuclear power plant (NPP). Through classifying and analyzing errors, a number of obtained data can be utilized to estimate current status and potential risks of software development and improve the quality of project. A method of error classification is proposed, which is applied to whole V and V life cycle, using a MW pressurized reactor project as an example. The purpose is to analyze errors discovered by V and V activities, and result in improvement of safety critical DCS application software. (authors)

  20. Modeling the critical safety functions status tree of a NPP using FPGA

    International Nuclear Information System (INIS)

    Farias, Marcos Santana; Oliveira, Mauro Vitor de; Jaime, Guilherme Dutra Gonzaga; Almeida, Jose Carlos Soares de; Augusto, Silas Cordeiro

    2013-01-01

    Field Programmable Gate Arrays (FPGAs) based systems and equipment are beginning to appear in new plants I and C applications, as well as in retrofits for operating plants, in particular for safety applications due to their capability to face the systems obsolescence since they are circuit independent. The circuits implemented can be portable to different FPGAs architectures. Moreover, they reduce complexity for regulatory approval as compared to conventional microprocessor-based systems. Critical safety function (CSF) is the most significant design concept for prioritize operator actions for NPP based on the potential threat to the three barriers (fuel cladding, primary coolant system boundary, and containment) and allows the operator to respond to these threats prior to event diagnosis. CSF has a hierarchical information structure that organizes the system variables affecting the plant safety in terms of goal-means relations. This paper describes the application of FPGA in the implementation of the CSFs status tree logic for a Westinghouse 3-loops NPP simulator. (author)

  1. Bases of updating of nuclear safety regulations for NPP in Romania

    International Nuclear Information System (INIS)

    Biro, Lucian; Serbanescu, Dan

    1999-01-01

    The paper presents the basic principles of reviewing and updating process of the regulatory environment pyramid. The main part of this review process refers to Cernavoda NPP Unit 2. However, there is an important impact on Cernavoda NPP Unit 1. The basic principles were defined in 1993/1994 when the licensing process for Unit 1 was resumed in order to be in accordance with the latest developments of Candu 600 worldwide and with the IAEA and NEA latest recommended documents and practices. After the licensing process for the Unit 1 was completed up to operation stage, CNCAN developed new updated regulations on nuclear safety and the regulatory pyramid in the framework of the RAMG PHARE project. CNCAN issued in 1996 the regulatory Policy for Unit 2, self-sustained and independent of future possible revisions of the regulations. The use of the concept of hierarchical systems, systematic review of safety criteria and objectives and margins along with the feedback from international and national experience on this topic ensured issuance of a reference document for future approach of nuclear safety in Romania. (authors)

  2. Safety culture in an RBMK perspective; Sostoyanie i perspektivy razvitiya sistemy povysheniya kul`tury bezopasnosti AEhS Rossii

    Energy Technology Data Exchange (ETDEWEB)

    Porokhin, V G [Rosenergoatom, Moscow (Russian Federation)

    1997-12-31

    The presentation discusses the following issues: state and perspectives of the development of the system on Russia NPP safety culture enhancement; steps of Rosehnergoatom on development of system on Russia NPP Russia NPP safety culture enhancement, qualitative and quantitative evaluation of the safety culture, the nearest perspectives on safety culture enhancement in Russia.

  3. KAERI software verification and validation guideline for developing safety-critical software in digital I and C system of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jang Yeol; Lee, Jang Soo; Eom, Heung Seop

    1997-07-01

    This technical report is to present V and V guideline development methodology for safety-critical software in NPP safety system. Therefore it is to present V and V guideline of planning phase for the NPP safety system in addition to critical safety items, for example, independence philosophy, software safety analysis concept, commercial off the shelf (COTS) software evaluation criteria, inter-relationships between other safety assurance organizations, including the concepts of existing industrial standard, IEEE Std-1012, IEEE Std-1059. This technical report includes scope of V and V guideline, guideline framework as part of acceptance criteria, V and V activities and task entrance as part of V and V activity and exit criteria, review and audit, testing and QA records of V and V material and configuration management, software verification and validation plan production etc., and safety-critical software V and V methodology. (author). 11 refs.

  4. KAERI software verification and validation guideline for developing safety-critical software in digital I and C system of NPP

    International Nuclear Information System (INIS)

    Kim, Jang Yeol; Lee, Jang Soo; Eom, Heung Seop.

    1997-07-01

    This technical report is to present V and V guideline development methodology for safety-critical software in NPP safety system. Therefore it is to present V and V guideline of planning phase for the NPP safety system in addition to critical safety items, for example, independence philosophy, software safety analysis concept, commercial off the shelf (COTS) software evaluation criteria, inter-relationships between other safety assurance organizations, including the concepts of existing industrial standard, IEEE Std-1012, IEEE Std-1059. This technical report includes scope of V and V guideline, guideline framework as part of acceptance criteria, V and V activities and task entrance as part of V and V activity and exit criteria, review and audit, testing and QA records of V and V material and configuration management, software verification and validation plan production etc., and safety-critical software V and V methodology. (author). 11 refs

  5. Utilization of a risk matrix based on Probabilistic Safety Analysis to improve nuclear safety in NPP

    International Nuclear Information System (INIS)

    Stubbe, Gerald

    2010-01-01

    PSA models. Those asymmetries are characterized by the fact that, for instance, a LOCA is considered to occur only on a certain primary loop. So, to take that weakness into account, we have decided to group the different redundant equipments of a function as one. This leads to some conservatism. For instance, if you have 3 LPSI pumps, you group those and consider as one. Uncertainties are also a threat for the PSA models. You must be sure that the model you use do not have 'cliff edge' effects. A complete sensitivity study can be enough to demonstrate that the uncertainties are bounded and understood. By use of the PSA models, we can give an importance factor to each equipment. This qualitative approach of quantitative data is expressed by two quantities: The Fussel Vessely (FV) and the Risk Increased Factor (RIF). The combination of those 2 factors can lead to the ranking of the equipment in three categories: Highly Safety Significant (HSS), Medium Safety Significant (MSS), Low Safety Significant (LSS). This ranking can be done globally, for all sequences, or initiator by initiator. The latter one is used to construct our matrix. It permits a larger comprehension of the phenomena following an initiator. And this is the goal of our risk matrix. A frequency of occurrence is assigned to each initiators, depending on generic data, or plant specific data for the more frequent ones. As an initiator occurs, you have a conditional probability to go to core damage. We call it the severity. We will assign a set of colour, representing the importance of the severity and the frequency of occurrence. This permits us to have a clear overview of what could happen to the plant. One of the main advantages of this matrix is to link, with a risk approach, equipments that are not linked together within the technical specifications of the plant. As example, in the NPP of GDF-Suez, we have a second level for the safeguards systems. In the design phase, it was to provide a protection

  6. Assessment of multi-version NPP I and C systems safety. Metric-based approach, technique and tool

    International Nuclear Information System (INIS)

    Kharchenko, Vyacheslav; Volkovoy, Andrey; Bakhmach, Eugenii; Siora, Alexander; Duzhyi, Vyacheslav

    2011-01-01

    The challenges related to problem of assessment of actual diversity level and evaluation of diversity-oriented NPP I and C systems safety are analyzed. There are risks of inaccurate assessment and problems of insufficient decreasing probability of CCFs. CCF probability of safety-critical systems may be essentially decreased due to application of several different types of diversity (multi-diversity). Different diversity types of FPGA-based NPP I and C systems, general approach and stages of diversity and safety assessment as a whole are described. Objectives of the report are: (a) analysis of the challenges caused by use of diversity approach in NPP I and C systems in context of FPGA and other modern technologies application; (b) development of multi-version NPP I and C systems assessment technique and tool based on check-list and metric-oriented approach; (c) case-study of the technique: assessment of multi-version FPGA-based NPP I and C developed by use of Radiy TM Platform. (author)

  7. Implementation of an indicator-based safety management system for the EnKK NPP's

    International Nuclear Information System (INIS)

    Bassing, G.; Nasellu, M.; Ritter, J.

    2004-01-01

    This presentation at the Eurosafe Berlin 2004 will give an overview of the activities on safety management of EnBW Kernkraft GmbH taken up with a notifiable event in August 2001 at KKP 2 nuclear site. After this event EnBW announced the development and introduction of an indicator-based safety management system (SMS) at all sites of its nuclear power plants. A SMS team with members from all NPP sites was built which had to analyse all processes based on the DIN EN ISO 9000 philosophy and to control them by indicators. The regulatory authorities and their experts would accompany this process in a suitable fashion and monitor and review it after its introduction. This presentation shows the process during the development of the system, the status of its introduction and the general involvement of the regulator. (orig.)

  8. Issue on NPP-I and C important to safety-Data Communication

    International Nuclear Information System (INIS)

    Koo, I. S.; Hong, S. B.; Cho, J. W.; Choi, Y. S.; Lee, J. C.

    2010-01-01

    1. Issue on CDV and FDIS of IEC61500 - Nuclear Power Plants - Instrumentation and control important to safety -Data communication - Activities on IEC TC45, SC45A/WGA3. 2. As issue the requirements for safety data communication which is essential part of digital I and C systems, the fundamental technology for IT based nuclear I and C is established. 3. Approval and circulation of IED61500 CDV and FDIS - Issue of the international standard, IEC 61500. 4. Issue one IEC61500, three interim documents, three presentations and five technical support to industry, and participation in IEC TC45 and SC45A plenary meeting and intermediate meeting on SC45A/WGA3. 5. Based on IEC61500, an new project on wireless technologyes application to NPP will be proceeded

  9. The safety of Ignalina NPP and ecological danger in public opinion of inhabitants of Daugavpils region

    International Nuclear Information System (INIS)

    Peipinia, O.

    1998-01-01

    Inquest of Daugavpils' region pointed to a big anxiety for ecological danger of Ignalina NPP by inhabitants and experts. Absolute majority of respondents (73-78% inhabitants and 68-82% experts) apprehend the NPP as very dangerous and dangerous. More than half of respondents apprehend the dangerous increased during last two-three years. It is because no one has a good reference about situation, because tragic al Chernobyl NPP burst was on. The anxiety increases if the respondent lives nearer of NPP. Inhabitants of Daugavpils and it's region wants the better reference about situation, about future of Ignalina NPP after 2010 year, about securities means in case of NPP burst. (author)

  10. The role of TSOs in the context of increasing demand for safety expertise - Expectations of the NPP operators

    International Nuclear Information System (INIS)

    Debes, M.

    2013-01-01

    This series of slides focuses on expectations from NPP operators concerning key points and role of TSOs (Technical Safety Organizations) related to -) skills and competence, with the need for broad systemic views regarding safety issues, -) the whole licensing and regulatory framework, for an efficient safety management in a competitiveness context (the requirements must be clear, reliable, stable, timely and internationally aligned), and -) the harmonization and the standardization in the licensing process to foster nuclear renaissance

  11. Safety assessment and regulatory strategy for NPP I and C modernization projects

    Energy Technology Data Exchange (ETDEWEB)

    Manners, S.; Blocquel, Ch

    1999-10-01

    IPSN is the technical support for the French nuclear safety authority (DSIN), but also acts independently. Through our participation at this IAEA meeting we wish to further our appreciation of the industry position for I and C modernization projects. We will present some of the concerns of the safety assessor and safety authority for such projects. We hope to share our experiences and views concerning current strategies for I and C modernization and licensing from. In our experience with NPP I and C programmes, the need for modification is most often not directly linked to safety. For our safety assessment we have to identify clearly and, as far as possible, categorize the safety relevance of the specified modifications and all safety impact in its implementation. Modernization can be simply for reasons of replacement of obsolete existing equipment or it can be linked to functional evolutions; safety functions may be directly or indirectly affected. The state of the art I and C solutions proposed by today's modernization programs have many benefits, but also pose a certain number of difficulties for the safety demonstration. On the implementation side, the safety assessment for a modernization project has to take into consideration specific issues compared with that for new plant. These include interface and compatibility with the existing installation, issues relating to 'on line' installation and commissioning, as well as operational issues concerning the changeover and trail periods. A further subject for discussion concerns how our regulatory requirements apply to modernization. We must as a minima comply with the requirements of the period. To what measure must we apply current or future (under development or for future reactor designs) standards? How can we tie in with requirements and legislation for new projects? Do we make a special case for back-fits? (authors)

  12. Safety assessment and regulatory strategy for NPP I and C modernization projects

    International Nuclear Information System (INIS)

    Manners, S.; Blocquel, Ch.

    1999-10-01

    IPSN is the technical support for the French nuclear safety authority (DSIN), but also acts independently. Through our participation at this IAEA meeting we wish to further our appreciation of the industry position for I and C modernization projects. We will present some of the concerns of the safety assessor and safety authority for such projects. We hope to share our experiences and views concerning current strategies for I and C modernization and licensing from. In our experience with NPP I and C programmes, the need for modification is most often not directly linked to safety. For our safety assessment we have to identify clearly and, as far as possible, categorize the safety relevance of the specified modifications and all safety impact in its implementation. Modernization can be simply for reasons of replacement of obsolete existing equipment or it can be linked to functional evolutions; safety functions may be directly or indirectly affected. The state of the art I and C solutions proposed by today's modernization programs have many benefits, but also pose a certain number of difficulties for the safety demonstration. On the implementation side, the safety assessment for a modernization project has to take into consideration specific issues compared with that for new plant. These include interface and compatibility with the existing installation, issues relating to 'on line' installation and commissioning, as well as operational issues concerning the changeover and trail periods. A further subject for discussion concerns how our regulatory requirements apply to modernization. We must as a minima comply with the requirements of the period. To what measure must we apply current or future (under development or for future reactor designs) standards? How can we tie in with requirements and legislation for new projects? Do we make a special case for back-fits? (authors)

  13. The management of safety within the completion of units 3. and 4. in NPP Mochovce

    International Nuclear Information System (INIS)

    Ocenas, M.

    2009-01-01

    The contribution defines the way of management of safety and protection of health and protection against the fire during completion of units 3. and 4. in NPP Mochovce. The creation of the management of safety raised from the legal and other requirements applied in the civil activities and mainly from the requirements for coordination of safety on the site. The whole system is methodically managed through the 5 basic documents so-called 'Building rules': 1) Safety-technical conditions of performance in SE, Inc. Bratislava (project MO34); 2) MO34/1/MNA-003.00-02 the elaboration of the integrated safety plan; 3) MO34/1/NA-003.00-03 risks analyse; 4) PNM3460100 Integrated safety plan for completion of Units 3.and 4 in NPP Mochovce; 5) Project of construction organization. The system itself of the management of the safety prevention must by provided by the supplier in compliance with the standard OHSAS 18 001 and ISO 14001, and in compliance with the requirement of the builder applied in the technical specification for the contract. Most of necessary related documentation of Safety plan for the site is usually part of the documentation for quality management of contractor. Therefore it is necessary to put the reference in each chapter of the Safety plan for site and the standpoint how to apply it for the particular specific situation. The coordinator is mandatory to manage the prevention for the part of safety of specified site and to coordinate the activity of suppliers during the works tie in this way, that the each contractor shall meet the safety requirements given by safety plan and also to identify and to eliminate the risks menaced in the particular work places or the suppliers each other. The builder established its own control system and for this purpose he settled the data base so-called 'The system of corrective and precautionary activities'. The system registers in its data base all discrepancies determined during the regular controls of the head workers

  14. Safety culture for engineering companies. Licensing and design bases for Cofrentes NPP

    International Nuclear Information System (INIS)

    Nhorte Gomez, M.D.

    1994-01-01

    Safety culture must be given higher priority by all organisations. It must not be considered a separate concept, attributable to just one particular organisation, or a single responsible party. It is important to apply this criterion throughout the different phases of a nuclear power plant project (design, construction, commissioning and operation) without becoming isolated or dissociated. Nevertheless, it is absolutely essential to apply and consider it during operation, so to ensure highest possible safety standards. Consideration must also be given to the interfaces and interconnections between the different parties involved in the project (Owner of the NPP, Main Engineering Company, Main Supplier, Regulatory Body, etc) to build a SAFETY CULTURE in a collective and effective way. In applying the safety culture, an engineering company emphasises the following concepts: - Personal dedication and sense of responsibility in all those involved in any activity related to the safety of Nuclear Power Plants. - Clearly defined and readily accessible areas of responsibility and channels of communication - Strict adherence to procedures - Internal review of activities (Design review) (Author)

  15. Problems and experience of ensuring nuclear safety in NPP spent fuel storage facilities in Russia

    International Nuclear Information System (INIS)

    Vnukov, Victor S.; Ryazanov, Boris G.

    2003-01-01

    The amount of Nuclear Power Plant (NPP) spent fuel in special storage facilities of Russia runs to more than 15000 tons and the annual growth is equal to about 850 tons. The storage facilities for spent nuclear fuel from the main nuclear reactors of Russia (RBMK-1000, VVER-1000, BN-600, EGP-6) were designed in the 60s - 70s. In the last years when the concept of closed fuel cycle and safety requirements had changed, the need was generated to have the nuclear storage facilities more crowded. First of all it is due to the necessity to increase the storage capacity because the RBMK-1000, VVER-1000, EGP-6 fuel is not reprocessed. So there comes the need for the facilities of a bigger capacity which meet the current safety requirements. The paper presents the results of studies of the most important nuclear safety issues, in particular: development of regulatory requirements; analysis of design-basis and beyond-the design-basis accidents (DBA and BDBA); computation code development and verification; justification of nuclear safety when water density goes down; the use of burn-up fraction values; the necessity and possibility to experimentally study the storage facility subcriticality; development of storage norms and rules for new types of fuel assemblies with mixed fuel and burnable poison. (author)

  16. Radiation safety of the helicopter air crow-liquidators of the Chernobyl NPP accident

    International Nuclear Information System (INIS)

    Davydov, B.I.; Ponomarenko, V.A.; Ushakov, I.B.

    1992-01-01

    Three aspects of the radiation safety of the helicopter air crew who took part in the elimination of the Chernobyl NPP accident after-effects during the period from the 27-th of April to July 1986 were considered: dosimetry, physical protection and radioprotector use. 1125 subjects from the flight and engineer personnel were irradiated: 6%-2.5 mSv and more, 32%-2.1-2.5 mSv and about 60% - up to 2.0 mSv. It is stated that the use of physical protection (local screening of a pilot seat and use of special antiradiation belts) makes irradiation doe 2-3 times less. In the intense conditions of flight activity the pilots preferred the radioprotector adrenaline to cystamin. Alimentary aids in the earlier and later periods after the accident influenced favourably somatic and psychic states. 16 refs.; 1 tab

  17. Main features of the unit 1-4 building complex, Kozloduy NPP in respect to seismic safety

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M; Boncheva, H; Stafanov, D [Central Laboratory for Seismic Mechanics and Earthquake Engineering, Bulgarian Academy of Sciences, Sofia (Bulgaria)

    1993-07-01

    The Units 1 and 2 of Kozloduy NPP were originally designed to resist a IV-V degree MSK earthquake. They have been subsequently upgraded for a VII degree earthquake. Since that structure basically do not meet the safety requirements to resist the new earthquake with a maximum acceleration of 0.2 g and very broad spectrum. The performed analyses are clearly pointing out that an upgrading for the new earthquake level is possible. The problems common for all the structures of Kozloduy NPP are summarized in this presentation.

  18. Main features of the unit 1-4 building complex, Kozloduy NPP in respect to seismic safety

    International Nuclear Information System (INIS)

    Kostov, M.; Boncheva, H.; Stafanov, D.

    1993-01-01

    The Units 1 and 2 of Kozloduy NPP were originally designed to resist a IV-V degree MSK earthquake. They have been subsequently upgraded for a VII degree earthquake. Since that structure basically do not meet the safety requirements to resist the new earthquake with a maximum acceleration of 0.2 g and very broad spectrum. The performed analyses are clearly pointing out that an upgrading for the new earthquake level is possible. The problems common for all the structures of Kozloduy NPP are summarized in this presentation

  19. Review of domestic and international experience on optimization of tests planning for safety related systems at NPP

    International Nuclear Information System (INIS)

    Skalozubov, V.I.; Komarov, Yu.A.; Kolykanov, V.N.; Kochneva, V.Yu.; Gablaya, T.V.

    2009-01-01

    There are represented the basic requirements of normative and operating documents on test periodicity of safety related systems at NPPs, sets out the theoretical methods of test optimization of the technical systems, and analyses foreign engineering methods for changing test periodicity of the NPP systems. Based on this review analyses further tasks are formulated for improvement of the methodical base of optimization of tests planning for safety related systems

  20. Quality Control Activities Related to Mechanical Maintenance of Safety Related Components at Krsko NPP

    International Nuclear Information System (INIS)

    Djakovic, D.

    2016-01-01

    For successful, safe and reliable operation of nuclear power plant, maintenance processes have to be systematically controlled and procedures for quality control of maintenance activities shall be established. This is requested by the quality assurance program, which shall provide control over activities affecting the quality of structures, systems, and components, considering their importance to safety. As a part of Quality and Nuclear Oversight Division (QNOD; SKV), the Quality Control Department (QC) provides quality control activities, which are deeply involved in maintenance processes at Krsko NPP, both on safety related and non-safety related (non-nuclear safety) components. QC activities on safety related components have to fulfil all requirements, which will enable the components to perform their intended safety functions. This paper describes quality control activities related to mechanical maintenance of safety related components at Krsko NPP and significant role of the Krsko plant QC Department in three particular maintenance cases connected with safety related components. In these three specific cases, the QC has confirmed its importance in compliance with quality assurance program and presented its significant added value in providing safe and reliable operation of the plant. The first maintenance activity was installation of nozzle check valves in the scope of a modification for improving regulation of spent fuel pit pumps. The QC Department performed receipt inspection of the valves. Using non-destructive examination methods and X-ray spectrometry, it was found out that the valve diffuser was made of improper material, which could cause progressive corrosion of the valve diffuser in borated water and consequently a loss of safety function of the valves followed by long-term consequences. The second one was the receipt inspection of containment ventilation fan coolers. The coolers were claimed and sent back to the supplier because the QC Department

  1. Electronuclear's safety culture assessment and enhancement program

    International Nuclear Information System (INIS)

    Selvatici, E.; Diaz-Francisco, J.M.; Diniz de Souza, V.

    2002-01-01

    The present paper describes the Eletronuclear's safety culture assessment and enhancement program. The program was launched by the company's top management one year after the creation of Eletronuclear in 1997, from the merging of two companies with different organizational cultures, the design and engineering company Nuclen and the nuclear directorate of the Utility Furnas, Operator of the Angra1 NPP. The program consisted of an assessment performed internally in 1999 with the support and advice of the IAEA. This assessment, performed with the help of a survey, pooled about 80% of the company's employees. The overall result of the assessment was that a satisfactory level of safety culture existed; however, a number of points with a considerable margin for improvement were also identified. These points were mostly related with behavioural matters such as motivation, stress in the workplace, view of mistakes, handling of conflicts, and last but not least a view by a considerable number of employees that a conflict between safety and production might exist. An Action Plan was established by the company managers to tackle these weak points. This Plan was issued as company guideline by the company's Directorate. The subsequent step was to detail and implement the different actions of the Plan, which is the phase that we are at present. In the detailing of the Action Plan, special care was taken to sum up efforts, avoiding duplication of work or competition with already existing programs. In this process it was identified that the company had a considerable number of initiatives directly related to organizational and safety culture improvement, already operational. These initiatives have been integrated in the detailed Action Plan. A new assessment, for checking the effectiveness of the undertaken actions, is planned for 2003. (author)

  2. The V-1 NPP and V-2 NPP upgrading

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities in the V-1 NPP and V-2 NPP upgrading as well as maintenance carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. The V-1 NPP applied the so called 'Small Backfitting Programme'covering 81 points of the Czechoslovak Atomic energy Commission Decree No 5/91. Continual upgrading continued after the Backfitting Programme completion with the Safety Report and following Nuclear Regulatory Authority of Slovak Republic (NRA SR) Decrees No 1/94 and 110/94 setting spheres and procedure for adopting and implementation of measures enabling the units to operate further on. Results of expert missions, analyses and assessments of components identified by Basic Engineering became the basis for the development of the Gradual Reconstruction Programme. The Programme outputs underwent economic and probabilistic assessing their contribution to nuclear safety. This process resulted in finalizing the Gradual Reconstruction Programme which started to be implemented in 1996 and will be completed in 1999. It is implemented by the REKON consortium and covers 17 areas including Instrumentation and Control, self-consumption emergency supply, leakage monitoring, emergency core cooling system, seismic reinforcement and radioactivity localisation. Both units will reach internationally acceptable safety standards for the remaining life-time period. The V-2 NPP Upgrading and Safety Enhancement Programme includes results of activities performed in the course of last years to define all important activities leading to enhancement of nuclear safety and performance reliability and effectiveness within the plant life-time period and to establish conditions for extending the life-time of these units for 40 years. The V-2 NPP Upgrading and Safety Enhancement Programme aims to assure safe operation with a probability of the core damages less than 10 -4 /reactor · year

  3. Probabilistic safety analysis for fire events for the NPP Isar 2

    International Nuclear Information System (INIS)

    Schmaltz, H.; Hristodulidis, A.

    2007-01-01

    The 'Probabilistic Safety Analysis for Fire Events' (Fire-PSA KKI2) for the NPP Isar 2 was performed in addition to the PSA for full power operation and considers all possible events which can be initiated due to a fire. The aim of the plant specific Fire-PSA was to perform a quantitative assessment of fire events during full power operation, which is state of the art. Based on simplistic assumptions referring to the fire induced failures, the influence of system- and component-failures on the frequency of the core damage states was analysed. The Fire-PSA considers events, which will result due to fire-induced failures of equipment on the one hand in a SCRAM and on the other hand in events, which will not have direct operational effects but because of the fire-induced failure of safety related installations the plant will be shut down as a precautionary measure. These events are considered because they may have a not negligible influence on the frequency of core damage states in case of failures during the plant shut down because of the reduced redundancy of safety related systems. (orig.)

  4. HSI for monitoring the critical safety functions status tree of a NPP

    International Nuclear Information System (INIS)

    Oliveira, Mauro Vitor de; Almeida, Jose Carlos Soares de; Augusto, Silas Cordeiro; Jaime, Guilherme Dutra Gonzaga

    2013-01-01

    Critical safety function (CSF) is the most significant design concept for prioritize operator actions based on the potential threat to the three barriers (fuel cladding, primary coolant system boundary, and containment) and allows the operator to respond to these threats prior to event diagnosis. CSF has a hierarchical information structure that organizes the system variables affecting the plant safety in terms of goal-means relations. It is important that the operator should be aware of various success paths associated with each CSF in order to respond to unanticipated system failures quickly. When an emergency occurs in NPPs, the operator should monitor CSFs periodically and identify possible success paths as necessary, and try to stabilize or safely shut down the plant using emergency operating procedure (EOP) that includes steps to check the CSFs. This implies that safety function status check may become a cognitively burdensome task that needs to be supported by proper information display. The advanced human-system interface (HSI) in nuclear power plants provides an information environment that supports the operators' burdensome cognitive tasks. This paper describes a CSFs interface design for supporting the operator's tasks to monitor and identify the associated success path for Westinghouse 3-loops NPP. (author)

  5. Design safety improvements of Kozloduy NPP to meet the modern safety requirements towards the old generation PWR

    International Nuclear Information System (INIS)

    Hinovski, M.P.; Sabinov, S.

    2001-01-01

    Activities related to safety improvement of Kozloduy NPP units, started at the end of 1970s included seismic resistance upgrading, fire safety improvement, reliable heat final absorber etc. During the last 10 years the approach was systematized and improved. Units 1 to 4 are of great interest; therefore here we will discuss these units only. As a result of studies and analyses performed at the end of the 1980s and the beginning of the 1990s, problems related to the safety were identified and complex of technical measures was developed and planned. A considerable part of these measures has already been implemented, and the rest will be performed during the next years. Activities were performed by stages, and at the moment the last stage is under way. It shall be finished by the year 2003. The number of the measures is quite large to describe them here in full scope -- during the first stage of the safety program (1991-1993) were developed and analyzed more than 4200 documents and more than 160 measures were executed. During the second and third stages more than 300 important improvements were realized. In the frame of the program, financed by EBRD, 10 new systems with great importance were implemented and 8 systems were significantly modified. The main measures are described below. (author)

  6. Nuclear safety in Slovak Republic. Regulatory aspects of NPP nuclear safety

    International Nuclear Information System (INIS)

    Lipar, M.

    1999-01-01

    Regulatory Authority (UJD) is appointed by the Slovak Republic National Council as an Executive Authority for nuclear safety supervision. Nuclear safety legislation, organisation and resources of UJD, its role and responsibilities are described together with its inspection and licensing functions and International cooperation concerning improvements of safety effectiveness. Achievements of UJD are listed in detail

  7. Nuclear Fuel in Cofrentes NPP

    International Nuclear Information System (INIS)

    2002-01-01

    Fuel is an essential in the nuclear power generating business because of its direct implications on safety, generating costs and the operating conditions and limitations of the facility. Fuel management in Cofrentes NPP has been targeted at optimized operation, enhanced reliability and the search for an in-depth knowledge of the design and licensing processes that will provide Iberdrola,as the responsible operator, with access to independent control of safety aspects related to fuel and free access to manufacturing markets. (Author)

  8. The regulatory requirements, design bases, researches and assessments in the field of Ukrainian NPP's seismic safety

    International Nuclear Information System (INIS)

    Mykolaychuk, O.; Mayboroda, O.; Krytskyy, V.; Karnaukhov, O.

    2001-01-01

    State Nuclear Regulatory Authority of Ukraine (SNRA) pays large attention to problem of nuclear installations seismic stability. As a result the seismic design regulatory guides is revised, additional seismic researches of NPP sites are conducted, seismic reassessment of NPP designs were begun. The experts involved address all seismic related factors under close contact with the staff of NPP, design institutes and research organizations. This document takes stock on the situation and the research programs. (author)

  9. The problem of maintenance of strength, lifetime and safety of the structural components operational NPP from items of a system approach

    International Nuclear Information System (INIS)

    Getman, A.F.

    2005-01-01

    The strength of the structural components and pipelines NPP determines largely their lifetime, radiation and nuclear safety. As shows world experience of operation NPP, in a series of cases during operation the local damages, breaks and destructions of elements of the equipment and pipelines emerge. These cases force maintaining organizations to execute on NPP large volumes of the control, repair and modernizing of the equipment and pipelines. These activities require large material inputs (net less than 50% of the cost of operation NPP), however not always are effective and in a series of cases do not allow to reach required of level reliability and safety. The reason of such condition of a problem is that the measures executed on NPP, developed on stretch of long time and under influence of the diverse factors. Besides until now there is no unified scientific methodology of a case study of maintenance of strength, lifetime and safety of the operational equipment and pipelines NPP. The application of a system approach to a problem of maintenance of strength, lifetime and safety of the equipment and pipelines operational NPP allows to consider a problem as a whole, from unified items, to define the most fast and effective paths it of the solution and to decide problems in as much as possible short times at minimum material inputs. The experience of practical application of a system approach at the solution of separate problems of operation NPP has allowed to develop a lot of effective new techniques, technologies and means, which application allows essentially to increase reliability and gamma percentage lifetime of elements NPP, to which they were applied. As an example in the report the outcomes of application of a system approach for maintenance of SG tubes integrity of reactors of a type WWER are adduced. The application of a system approach for all elements of the equipment and pipelines NPP will allow to increase it reliability and safety not less, than on 10

  10. Safety assessment of Olkiluoto NPP units 1 and 2. Decision of the Radiation and Nuclear Safety Authority regarding the periodic safety review of the Olkiluoto NPP

    International Nuclear Information System (INIS)

    2010-02-01

    In this safety assessment the Radiation and Nuclear Safety Authority (STUK) has evaluated the safety of the Olkiluoto Nuclear Power Plant units 1 and 2 in connection with the periodic safety review. This safety assessment provides a summary of the reviews, inspections and continuous oversight carried out by STUK. The issues addressed in the assessment and the related evaluation criteria are set forth in the nuclear energy and radiation safety legislation and the regulations issued thereunder. The provisions of the Nuclear Energy Act concerning the safe use of nuclear energy, security and emergency preparedness arrangements, and waste management are specified in more detail in the Government Decrees and Regulatory Guides issued by STUK. Based on the assessment, STUK consideres that the Olkiluoto Nuclear Power Plant units 1 and 2 meet the set safety requirements for operational nuclear power plants, the emergency preparedness arrangements are sufficient and the necessary control to prevent the proliferation of nuclear weapons has been appropriately arranged. The physical protection of the Olkiluoto nuclear power plant is not yet completely in compliance with the requirements of Government Decree 734/2008, which came into force in December 2008. Further requirements concerning this issue based also on the principle of continuous improvement were included in the decision relating to the periodic safety review. The safety of the Olkiluoto nuclear power plant was assessed in compliance with the Government Decree on the Safety of Nuclear Power Plants (733/2008), which came into force in 2008. The decree notes that existing nuclear power plants need not meet all the requirements set out for new plants. Most of the design bases pertaining to the Olkiluoto 1 and 2 nuclear power plant units were set in the 1970s. Substantial modernisations have been carried out at the Olkiluoto 1 and 2 nuclear power plant units since their commissioning to improve safety. This is in line with

  11. Analysis of Paks NPP Personnel Activity during Safety Related Event Sequences

    International Nuclear Information System (INIS)

    Bareith, A.; Hollo, Elod; Karsa, Z.; Nagy, S.

    1998-01-01

    Within the AGNES Project (Advanced Generic and New Evaluation of Safety) the Level-1 PSA model of the Paks NPP Unit 3 was developed in form of a detailed event tree/fault tree structure (53 initiating events, 580 event sequences, 6300 basic events are involved). This model gives a good basis for quantitative evaluation of potential consequences of actually occurred safety-related events, i.e. for precursor event studies. To make these studies possible and efficient, the current qualitative event analysis practice should be reviewed and a new additional quantitative analysis procedure and system should be developed and applied. The present paper gives an overview of the method outlined for both qualitative and quantitative analyses of the operator crew activity during off-normal situations. First, the operator performance experienced during past operational events is discussed. Sources of raw information, the qualitative evaluation process, the follow-up actions, as well as the documentation requirements are described. Second, the general concept of the proposed precursor event analysis is described. Types of modeled interactions and the considered performance influences are presented. The quantification of the potential consequences of the identified precursor events is based on the task-oriented, Level-1 PSA model of the plant unit. A precursor analysis system covering the evaluation of operator activities is now under development. Preliminary results gained during a case study evaluation of a past historical event are presented. (authors)

  12. Krsko NPP Quality Assurance Plan Application to Nuclear Safety Upgrade Projects (PCFV System and PAR System)

    International Nuclear Information System (INIS)

    Biscan, Romeo; Fifnja, Igor

    2014-01-01

    Nuklearna Elektrarna Krsko (NEK) has undertaken Nuclear Safety Upgrade Projects as a safety improvement driven by the lessons learned from the Fukushima-Daiichi Accident. Among other projects, new modification 1008-VA-L Passive Containment Filtered Vent (PCFV) System has been installed which acts as the last barrier minimizing the release of radioactive material into the environment in case of failure of all safety systems, and to insure containment integrity during beyond design basis accidents (BDBA). In addition, modification 1002-GH-L Severe Accident Hydrogen Control System (PAR) has been implemented to prevent and mitigate the consequences of explosive gas generation (hydrogen and carbon monoxide) in case of reactor core melting. To ensure containment integrity for all design basis accidents (DBA) and BDBA conditions, NEK has eliminated existing safety-related electrical recombiners, replaced them with two safety-related passive autocatalytic recombiners (PARs) and added 20 new PARs designed for the BDBA conditions. Krsko NPP Quality Assurance Plan has been applied to Nuclear Safety Upgrade Projects (PCFV System and PAR System) through the following activities: · Internal audit of modification process was performed. · Supplier audits were performed to evaluate QA program efficiency of the main design organization and engineering organizations. · Evaluation and approval of Suppliers were performed. · QA engineer was involved in the review and approval of 1008-VA-L and 1002-GH-L modification documentation (Conceptual Design Package, Design Modification Package, Installation Package, Field Design Change Request, Problem/Deficiency Report, and Final Documentation Package). · Purchasing documentation for modifications 1008-VA-L and 1002-GH-L (technical specifications, purchase orders) has been verified and approved by QA. · QA and QC engineers were involved in oversight of production and testing of the new 1008-VA-L and 1002-GH-L plant components.

  13. Feedback from peer review missions of the self-assessment of operational events reflecting safety culture (Leningrad, Smolemsk and Balakovo NPP`s, the Russian Federation)

    Energy Technology Data Exchange (ETDEWEB)

    Gantchev, T [Committee on the Use of Atomic Energy for Peaceful Purposes (Bulgaria)

    1997-10-01

    The presentation discusses the following issues: criteria for reporting events inside NPP; screening of the plant operational events; third safety culture indicator; list of the pending safety problems; significance of the safety problems for plant reliability; prioritization of the safety culture issues. Tabs.

  14. Feedback from peer review missions of the self-assessment of operational events reflecting safety culture (Leningrad, Smolemsk and Balakovo NPP's, the Russian Federation)

    International Nuclear Information System (INIS)

    Gantchev, T.

    1997-01-01

    The presentation discusses the following issues: criteria for reporting events inside NPP; screening of the plant operational events; third safety culture indicator; list of the pending safety problems; significance of the safety problems for plant reliability; prioritization of the safety culture issues. Tabs

  15. Experience of RIA safety analyses performance for NPP Temelin core arranged with TVSA-T fuel assemblies

    International Nuclear Information System (INIS)

    Kryukov, S.A.; Lizorkin, M.P.

    2010-01-01

    The contents of the presentation are as follows: 1. Definition of categories for initiating events; 2. Acceptance criteria for safety assessment; 3. Main aspects of safety assessment methodology; 4. Main stages of calculation analysis; 5. Interface with other parts of the core design; 6. Codes used for calculation; 6.1 Main performances of code package TIGR-1; 6.2 Main performances of code BIPR-7A; 7. TIGR-1 accounting of design margins in calculation of fuel rod powers; 8. Peculiar features of Instrumentation and Control System for Temelin NPP; 9. Calculations; 10. Checklist of margin data important for reload safety assessment. (P.A.)

  16. Estimation of the loss of Offsite power frequency for the probabilistic safety assessment of the Juragua NPP

    International Nuclear Information System (INIS)

    Vilaragut Llanes, J.J.; Valhuerdi Debesa, C.

    1996-01-01

    The loss offsite power is defined as the interruption of the preferred power supply to the essential and non essential switchgear buses necessitating or resulting in the use of emergency AC power supply. Because many safety system required for reactor core decay heat removal and containment heat removal depend on AC power, a loss of offsite power, if emergency power supply (diesel generators) fails, could be severe accidents The purpose of this work was to determine, for the Probabilistic Safety Assessment of the Juragua NPP, the causes, frequency and duration relationships of the loss of offsite power. A description is presented of the different factor that determine the occurrence of this event and the characteristics for the Juragua NPP

  17. Engineering safety review mission Krsko NPP external events PSA. Ljubljana, Slovenia 19-23 February 1996. Final report

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Smith, P.

    1996-01-01

    Within the scope of the TC Project RER/9/035, a review mission visited Ljubljana, Slovenia, 19-23 February 1996. Two outside experts, Messrs. R.J. Budnitz (USA) and Paul Smith (USA), as well as a staff member, A. Guerpinar (ESS-NSNI) took part in the review. The purpose of the mission was to assist the Slovenian Nuclear Safety Administration to review the external events PSA prepared by Krsko NPP consultants Westinghouse Energy Systems Europe and EQE International. Another seismic safety review was performed concurrently in Ljubljana involving the investigations in relation to the tectonic stability and reassessment of the design basis ground motion characterization for the Krsko NPP site

  18. Emergency preparedness at Ignalina NPP

    International Nuclear Information System (INIS)

    Kairys, A.

    1998-01-01

    Brief review of Ignalina NPP safety upgrading and personnel preparedness to act in cases of accidents is presented. Though great activities are performed in enhancing the plant operation safety, the Ignalina NPP management pays a lot of attention to preparedness for emergency elimination and take measures to stop emergency spreading. A new Ignalina NPP emergency preparedness plan was drawn up and became operational. It is the main document to carry out organizational, technical, medical, evacuation and other activities to protect plant personnel, population, the plant and the environment from accident consequences. Great assistance was rendered by Swedish experts in drawing this new emergency preparedness plan. The plan consists of 3 parts: general part, operative part and appendixes. The plan is applied to the Ignalina NPP personnel, Special and Fire Brigade and also to other contractor organizations personnel carrying out works at Ignalina NPP. There are set the following emergency classes: incident, emergency situation, alert, local emergency, general emergency. Separate intervention level corresponds to each emergency class. Overview of personnel training to act in case of an emergency is also presented

  19. Leadership and Safety Management: Regulatory Initiatives for Enhancing Nuclear Safety in the Republic of Korea

    International Nuclear Information System (INIS)

    Yun, C.H.; Park, Y.W.; Choi, K.S.

    2010-01-01

    Since the construction of the first nuclear power plant (NPP) in the Republic of Korea in 1978, a high level of nuclear safety has continued to be maintained. This has been the important basis on which the continuous construction of NPPs has been possible in the country. To date, regulatory initiatives, leaderships and strategies adopting well harmonized regulatory systems and practices of advanced countries have contributed to improving the effectiveness and efficiency of safety regulation and further enhancing nuclear safety. The outcomes have resulted in a high level of safety and performance of Korean NPPs, attributing largely to the safety promotion policy. Recently, with the support of the Korean Ministry of Education, Science and Technology (MEST), the Korea Institute of Nuclear Safety (KINS) established the International Nuclear Safety School and created a Nuclear Safety Master's Degree Programme. Further, it developed multilateral and bilateral cooperation with other agencies to promote global nuclear safety, with the aim of providing knowledge and training to new entrant countries in establishing the safety infrastructure necessary for ensuring an acceptable level of nuclear safety. (author)

  20. Safety requirement of the nuclear power plants, after TMI-2 accident and their possible implementation on Bushehr NPP

    International Nuclear Information System (INIS)

    Mirhabibi, N.; Tochai, M.T.M.; Ashrafi, A.; Farnoudi, E.

    1985-01-01

    Based on the lessons learned from the TMI-2 accident and other research and developments, many improvements have been required for the design, manufacturing and operation of nuclear power plants in recent years. These requirements have already been implemented to the plants in operation and considered as new safety requirements for new plants. In the present paper these requirements and their possible implementation on Bushehr NPP are discussed. (Author)

  1. Research for enhancing reactor safety

    International Nuclear Information System (INIS)

    1989-05-01

    Recent research for enhanced reactor safety covers extensive and numerous experiments and computed modelling activities designed to verify and to improve existing design requirements. The lectures presented at the meeting report GRS research results and the current status of reactor safety research in France. The GRS experts present results concerning expert systems and their perspectives in safety engineering, large-scale experiments and their significance in the development and verification of computer codes for thermohydraulic modelling of safety-related incidents, the advanced system code ATHLET for analysis of thermohydraulic processes of incidents, the analysis simulator which is a tool for fast evaluation of accident management measures, and investigations into event sequences and the required preventive emergency measures within the German Risk Study. (DG) [de

  2. Analytical tool for the periodic safety analysis of NPP according to the PSA guideline. Vol. 1

    International Nuclear Information System (INIS)

    Balfanz, H.P.; Boehme, E.; Musekamp, W.; Hussels, U.; Becker, G.; Behr, H.; Luettgert, H.

    1994-01-01

    The SAIS (Safety Analysis and Informationssystem) Programme System is based on an integrated data base, which consists of a plant-data and a PSA related data part. Using SAIS analyses can be performed by special tools, which are connected directly to the data base. Two main editors, RISA+ and DEDIT, are used for data base management. The access to the data base is done via different types of pages, which are displayed on a displayed on a computer screen. The pages are called data sheets. Sets of input and output data sheets were implemented, such as system or component data sheets, fault trees or event trees. All input information, models and results needed for updated results of PSA (Living PSA) can be stored in the SAIS. The programme system contains the editor KVIEW which guarantees consistency of the stored data, e.g. with respect to names and codes of components and events. The information contained in the data base are called in by a standardized users guide programme, called Page Editor. (Brunsbuettel on reference NPP). (orig./HP) [de

  3. Management of fire and industrial safety - challenges during commissioning of a NPP

    International Nuclear Information System (INIS)

    Maiti, Subhaschandra; Mohan, Nalini; Ghadge, S.G.; Bajaj, S.S.

    2006-01-01

    Construction and commissioning period of NPP are reduced world over drastically by stringent schedule for financial and economic reasons. For meeting the schedule, commissioning of components and systems are started immediate after installation, while construction activities are continued in parallel at the same place. Parallel activities' and 'Time Constraint' have brought new challenges to 'Management of Fire and Industrial Safely' during commissioning. An innovative approach was used during such phase of commissioning of TAPP-3 and 4. This paper outlines challenges encountered during this phase and special approach and measures used to meet those challenges. This paper also outlines problems encountered during implementation of these measures and subsequent change in approach to ensure smooth and safe execution of activities. Primarily, challenges were conflicting requirements by various agencies to carryout commissioning in parallel with construction activities concurrently. Main challenges were related to fall hazard, chemical hazard, fire hazard, electrical safety, work in confined space, housekeeping problem. Moreover it was within exclusion zone of another operating plant, which added one more dimension to those challenges. Conventional Safely management approach was little short to resolve these challenges. Such challenges were envisaged; analyzed and innovative measures were arrived at. Along with conventional safely analysis like Job Safely or Hazard Analysis (JSA or JHA), Accident Analysis, Accident Trend Analysis, Fire Hazard Analysis (FHA), innovative method like 'Area-Job- Hazard Charting' and 'Ratio Analysis' were used to understand activity dependent time varying hazard scenarios. Based on this analysis, decision were taken to change various existing elements of safely management like safety organizations; standard operating procedures (SOP); emergency operating procedures (EOP); resource allocation, planning and scheduling; safely training; safely

  4. Safety Culture Enhancement Project. Final Report. A Field Study on Approaches to Enhancement of Safety Culture

    International Nuclear Information System (INIS)

    Lowe, Andrew; Hayward, Brent

    2006-08-01

    Safety Culture Perceptions Questionnaire conducted with site managers to access their opinions about the adequacy of the local safety culture; - a framework of safety-related competencies for managers, representing desirable actions for leading and promoting a positive safety culture; - results of an evaluation survey completed by participants at the conclusion of the Management Workshops to assess the utility of this activity. Section 4 of the report, Recommendations for Future Action, highlights nine proposed activities that could be undertaken to build on the outcomes from this project, to support the enhancement of safety culture within the Swedish nuclear industry in the longer term. Specifically, these recommendations propose actions to: 1. Introduce formal processes to ensure the ongoing development of safety related competencies amongst industry managers. 2. Strengthen the resources, contribution, value and profile of Man Technology Organisation (MTO) expertise within nuclear industry sites, in order to promote a better understanding of human performance issues, enhance error management and accident prevention capabilities. 3. Identify ways to embed existing positive safety culture attributes, in an environment of considerable workforce changes resulting from increasing use of contractors and (expected) retirements amongst an ageing industry employee population. 4. Standardise and improve aspects of incident and accident investigation processes and analysis methodologies currently used, to improve information sharing and optimise learning. 5. Implement harmonised MTO / human factors awareness training programs at appropriate levels for all nuclear industry personnel. 6. Formalise the application of applied teamwork training (as per the principles of Crew Resource Management training in aviation) for NPP Control Room Operators, Maintenance workers and other employees working in safety-critical teams. 7. Increase the use of simulation training to enhance non

  5. Safety Culture Enhancement Project. Final Report. A Field Study on Approaches to Enhancement of Safety Culture

    Energy Technology Data Exchange (ETDEWEB)

    Lowe, Andrew; Hayward, Brent (Dedale Asia Pacific, Albert Park VIC 3206 (Australia))

    2006-08-15

    results of the Safety Culture Perceptions Questionnaire conducted with site managers to access their opinions about the adequacy of the local safety culture; - a framework of safety-related competencies for managers, representing desirable actions for leading and promoting a positive safety culture; - results of an evaluation survey completed by participants at the conclusion of the Management Workshops to assess the utility of this activity. Section 4 of the report, Recommendations for Future Action, highlights nine proposed activities that could be undertaken to build on the outcomes from this project, to support the enhancement of safety culture within the Swedish nuclear industry in the longer term. Specifically, these recommendations propose actions to: 1. Introduce formal processes to ensure the ongoing development of safety related competencies amongst industry managers. 2. Strengthen the resources, contribution, value and profile of Man Technology Organisation (MTO) expertise within nuclear industry sites, in order to promote a better understanding of human performance issues, enhance error management and accident prevention capabilities. 3. Identify ways to embed existing positive safety culture attributes, in an environment of considerable workforce changes resulting from increasing use of contractors and (expected) retirements amongst an ageing industry employee population. 4. Standardise and improve aspects of incident and accident investigation processes and analysis methodologies currently used, to improve information sharing and optimise learning. 5. Implement harmonised MTO / human factors awareness training programs at appropriate levels for all nuclear industry personnel. 6. Formalise the application of applied teamwork training (as per the principles of Crew Resource Management training in aviation) for NPP Control Room Operators, Maintenance workers and other employees working in safety-critical teams. 7. Increase the use of simulation training to

  6. Ignalina NPP its environment, safety and future, prospects of the energetic, ethnic and cultural situation: expert evaluation

    International Nuclear Information System (INIS)

    Morkunas, Z. V.; Ciuzas, A.; Jonaitis, V.; Sutiniene, I.

    1995-01-01

    According to the tasks defined in the 'Atomic Energy and the Environment' program an expert evaluative survey was done for the first time in Lithuania concerning the Ignalina NPP and its consequences and perspectives according to the concept which was prepared. The results of survey analysis, done by Lithuanian experts, are presented. Investigation involved these problems: evolution of the technical state safety, use and prospects of the nuclear power plant; evaluation of the activities of governmental and social institutions in connection with the nuclear power plant; Ignalina NPP and the environment; the effect of the nuclear power plant on agricultural activities and development; evolution of the ethnic and cultural situation; conclusions and recommendations for regulations of those areas. (author). 2 refs., 11 figs

  7. The World Bank mission in Bulgaria with the objective of estimating the possibility to increase the safety of the Kozloduy NPP

    International Nuclear Information System (INIS)

    Laaksonen, J.

    1992-01-01

    This is a report of the World Bank mission in Bulgaria held from 22 March to 31 April 1992. The author considers the present state of the safety level of all Kozloduy NPP reactors with emphasis of the units 1-4 which are of the WWER-440/230 type. Three main factors are discussed in detail: 1) design inherent safety; 2) equipment protection from accidental events connected with the equipment quality; 3) quality and culture of operation and quantitative evaluation of the safety. The areas requiring special attention in view of the safety enhancing are: safety goal policy; creation of national safety culture; planning and implementation of the improvement programme; assuring the equipment reliability; updating and protection the existing systems; furnishing the plant with new systems. A review of all on-going and proposed projects of the national authorities, IAEA, EC, WANO and BEQE is made. Some remarks are given also about the WWER-1000 type units 5 and 6. The estimated cost of all necessary measures for equipment improvement is $100-200 million and of the annual investment after the main reconstruction and innovation of the reactors - about $10 million which should be included in the normal operational expenses of the plant. A new approach to the responsibility role of the Bulgarian specialists should be created. A new and well grounded proposal for modifications in the equipment should be made from the National Electrical Company including economical evaluations. Another expert assessment will be requested from the foreign banks before financing

  8. Belene NPP

    International Nuclear Information System (INIS)

    Tsvetanov, P.

    1990-01-01

    The book presents the main results of the studies of the Bulgarian Academy of Sciences (BAS) on the construction of a new nuclear power plant at Belene on the Danube river. The programme of the studies comprises five areas: the socio-economic and energy development and the necessity of the commissioning; a technical project and design level of the equipment (safety, radioactivity control, waste disposal and economic efficiency of the power plant); the seismic properties of the construction site; the corresponding risk and design features of the plant; the ecological impacts of the NPP and public opinion. The studies in the different areas have been carried out by independent teams, fully responsible for the formulated topical conclusions. The general opinion of the BAS voiced in the book is that the construction of Belene NPP is not sufficiently substantiated and is considered unacceptable. 94 refs., 53 fig., 56 tabs. (R.Ts.)

  9. Application of safety standards and rules in the Shelter Implementation Plan at the destroyed power unit of Chernobyl NPP

    International Nuclear Information System (INIS)

    Berthold, A.; Bogorinski, P.; Bykov, V.; Redko, V.; Erickson, L.; Kadkin, Ye.; Kondratiev, S.; Simonov, I.; Smyshliaieva, S.; Yesipenko, Yu.

    2002-01-01

    This report deals with the application of safety standards and rules to the Shelter Implementation Plan (SIP) measures. Since 1998 this plan is being implemented at the Chornobyl NPP destroyed unit (which is now known as the Shelter). It includes a set of various tasks whose performance will help partially achieve the established safety objectives. The Regulatory Authority should establish for the Shelter safety goals, principles, and criteria in general, while the Operator of the Shelter is free to independently select the optimum method for their implementation. The Operator of the Shelter must demonstrate (in safety analysis report) that established safety goals are achieved and safety principles and criteria are met. Safety goals, principles, and criteria established for radioactive waste management are reasonable to apply in measures provided for by SIP. However, due to the unique nature of the Shelter, some criteria should not be applied directly and in full scope. Norms and rules on radiation protection should be applied in full scope. The specifics of radiation protection during each Shelter-related activity are considered individually. Safety standards and rules related to technical aspects are reasonable only as a basis. Effective resolution of specific technical issues associated with safety assurance is achieved through interaction between the Operator and the Regulatory Authority during design of SIP structures and systems. Hence, effectiveness of the licensing process plays an important role in the success of the SIP.(author)

  10. The comparative analysis of safety and economic competitiveness of the advanced high-power reactor projects of NPP

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Makhanov, U.M.; Philimonova, R.A.; Kichutkina, E.G.

    2002-01-01

    The comparative analysis results of the safety and economic competitiveness of the seven advanced large sized reactors projects (900 MW and more) are submitted in that report: EPR, Frameatome France, Siemens Germany; EP-1000 Westinghouse, USA and Genesi Italy; Candu 9, Atomic Energy of Canada Ltd; System 80 +, ABB, USA; KNGR, group NSSS Engineering and Development, Korea Power Engineering Company, Inc; APWR, Electric Power company, Japan Atomic Power Company, Mitsubishi Heavy Industries, Westinghouse Electric; and WWER-1000 (V-392), Atomenergoproject/Gidropress Russian Federation. According to the economic competitiveness of listed compared power reactors the 14 criteria of safety have been accepted. These criteria: 1. Features of the barrier system of 'defence-in-depth'. 2. The self-security of a reactor under increase of power and reactivity of a reactor, decrease of the expense and phase transformations of the reactor core coolant (presence of negative feedbacks). 3. Presence of the reactor shutdown systems responding principles of a variety, independence and reservation. Presence of the passive means of initiation and operation of the emergency protection. 4. The emergency cooling of the core of reactor. A presence of the passive means of cooling. Presence of the water reservation for the water supply of the different safety systems. 5. The emergency electrical supply, its reliability and degree of reservation. 6. The prevention measures of the heavy accident with the melt core. The decrease of the heavy accident probability. 7. The account of the heavy accident under development of the levels of protection. 8. The protection levels of NPP, the technological criteria of efficiency of the each safety barriers and the limiting radiation criteria for the each level of protection , in particular for the design-basis and beyond-design-basis accidents. 9. The measures for reduction of the heavy accident consequences. The management by the beyond

  11. International experts conclude IAEA peer review of Iran's safety regulation of Bushehr NPP

    International Nuclear Information System (INIS)

    2010-01-01

    Full text: An international team of nuclear safety experts today completed an IAEA mission to review the effectiveness of Iran's safety regulation of its first nuclear power plant and to identify possible improvements before the plant begins operation. Upon invitation of the Islamic Republic of Iran, the International Atomic Energy Agency (IAEA) assembled a team of senior regulators from seven Member States for an Integrated Regulatory Review Service (IRRS) mission. The scope of the mission was limited to the safety regulation of Bushehr nuclear power plant (BNPP-1). The IRRS review took place from 20 February to 2 March at the INRA offices in Tehran and included a technical visit to the BNPP-1 site. The mission was an objective peer review based on IAEA safety standards, and was neither an inspection, nor an audit. Ms. Olena Mykolaichuk, IRRS Team Leader and Head of the State Nuclear Regulatory Committee of Ukraine, commended her INRA counterparts: 'The regulatory work performed on the Bushehr construction and in preparation for commissioning has demonstrated significant progress of INRA as a nuclear regulatory authority,' she said. Philippe Jamet, Director of the IAEA's Nuclear Installation Safety Division, added: 'Through this IRRS mission, both Iran and the international experts contribute to the enhancement of nuclear safety and worldwide experience sharing.' In the course of its review the IRRS team identified the following strengths: - INRA has a dedicated, conscientious staff, demonstrating clear commitments to further improvements. - INRA clearly recognizes the value of peer reviews and international cooperation regarding nuclear safety. - Despite a shortage of staff, INRA demonstrated strong leadership while performing both review and assessment and inspection tasks during the BNPP-1 construction and pre-commissioning. - INRA has developed an excellent computerized documentation control system. Recommendations and suggestions to improve INRA's regulatory

  12. Regulatory requirements for radiation safety in the design of a new Finish NPP

    Energy Technology Data Exchange (ETDEWEB)

    Alm-Lytz, Kirsi; Vilkamo, Olli [Radiation and Nuclear Safety Authority, STUK, PO Box 14, Laippatie 4, 00881 Helsinki (Finland)

    2004-07-01

    There are two operating nuclear power plants in Finland, two BWR units at Olkiluoto site and two PWR units at Loviisa site. These reactors were commissioned between 1977 and 1981. The total electricity capacity in Finland is about 15 GW. In 2003, nuclear power plants generated one fourth of Finland's electricity. Despite of the diversity of the electricity generation methods, Finland is highly dependent on imported energy. Electricity consumption is estimated to increase and the demand for extra capacity has been estimated at about 2500-3000 MW by 2010. It should also be taken into account that a considerable proportion of the production capacity constructed in the 1970's must be replaced with production capacity of new power plants in the near future. In practice, the climate politics commitments made by Finland exclude coal power. Therefore, the capacity can be increased significantly only by natural gas, nuclear power and biofuels. The paper presents the following issues: Licensing a new nuclear power plant in Finland; FIN5 Project at STUK; Work planning and a tool for requirement management; Radiation safety related YVL guides; Collective dose target; On-site habitability during accident situation. Habitability was evaluated on the basis of the calculated dose rate levels, the occupancy times and the dose limits. Radiation hazard was classified into three parts, i.e., possible direct radiation from the containment, air contamination and systems carrying radioactive air or water. The results showed that direct radiation from the containment is generally adequately shielded but penetrations and hatches have to be separately analysed and the radiation dose levels near them are usually rather high. Skyshine radiation from the reactor containment is a special feature at the Loviisa NPP and the nearby area outside the buildings might have very limited access for the first hours after the accident. The skyshine effect is not usually relevant hazard in

  13. Enhancement and evaluation of an algorithm for atmospheric profiling continuity from Aqua to Suomi-NPP

    Science.gov (United States)

    Lipton, A.; Moncet, J. L.; Payne, V.; Lynch, R.; Polonsky, I. N.

    2017-12-01

    We will present recent results from an algorithm for producing climate-quality atmospheric profiling earth system data records (ESDRs) for application to data from hyperspectral sounding instruments, including the Atmospheric InfraRed Sounder (AIRS) on EOS Aqua and the Cross-track Infrared Sounder (CrIS) on Suomi-NPP, along with their companion microwave sounders, AMSU and ATMS, respectively. The ESDR algorithm uses an optimal estimation approach and the implementation has a flexible, modular software structure to support experimentation and collaboration. Data record continuity benefits from the fact that the same algorithm can be applied to different sensors, simply by providing suitable configuration and data files. Developments to be presented include the impact of a radiance-based pre-classification method for the atmospheric background. In addition to improving retrieval performance, pre-classification has the potential to reduce the sensitivity of the retrievals to the climatological data from which the background estimate and its error covariance are derived. We will also discuss evaluation of a method for mitigating the effect of clouds on the radiances, and enhancements of the radiative transfer forward model.

  14. Post Fukushima safety enhancements in Indian PHWRS

    International Nuclear Information System (INIS)

    Ramasomayajulu, M.; Khot, Pankaj; Chauhan, Ashok

    2016-01-01

    Fukushima event was reviewed in Nuclear Power Corporation of India (NPCIL) and based on these reviews, safety enhancements were identified for Indian PHWRs. Safety enhancements such as additional emergency power sources, enhanced onsite water inventories, external water injection arrangements (Hook up points), measures related to hydrogen management, containment venting provision, seismic trip, mobile pumps, onsite emergency support Centre. These safety enhancements were reviewed by the regulatory body (Atomic Energy Regulatory Board, AERB) and were approved for implementation. Most of these are either implemented or in the advance stage of implementation. The paper elaborates above safety enhancements implemented post Fukushima accident; and preparedness to use these provisions. (author)

  15. Operational Readiness Verification, Phase 3. A Field Study at a Swedish NPP during a Productive Outage (Safety-train Outage)

    International Nuclear Information System (INIS)

    Hollnagel, Erik; Gauthereau, Vincent; Persson, Bodil

    2004-01-01

    This report describes the results from Phase III of a study on Operational Readiness Verification (ORV) that was carried out from December 2002 to November 2003. The work comprised a field study of ORV activities at a Swedish NPP during a planned productive outage [subavstaellning], which allowed empirical work to be conducted in an appropriate environment with good accessibility to technical staff. One conclusion from Phase I of this project was the need to look more closely at the differences between three levels or types of tests that occur in ORV: object (component) test, system level test and (safety) function test, and to analyse the different steps of testing in order to understand the nontrivial relations between tests and safety. A second conclusion was the need to take a closer look at the organisation's ability to improvise in the sense of adjusting pre-defined plans to the actual conditions under which they are to be carried out. Phase II of the project found that although all three types of test occurred, they were rather used according to need rather than to a predefined arrangement or procedure. The complexity of ORV could be understood and described by using the concepts of Community of Practice, embedding, and Efficiency-Thoroughness Trade-Off. In addition, organisation and the different communities of practice improvise by adjusting pre-defined plans or work orders to the existing conditions. Such improvisations take place both on the levels of individual actions, on the level of communities of practice, and on the organisational level. The ability to improvise is practically a necessity for work to be carried out, but is also a potential risk. Phase III of the project studied how tasks are adapted relative to the different types of embedding and the degree of correspondence between nominal and actual ORV. It also looked further at the different Communities of Practice that are part of maintenance and ORV, focusing on the coordination and

  16. Operational Readiness Verification, Phase 3. A Field Study at a Swedish NPP during a Productive Outage (Safety-train Outage)

    Energy Technology Data Exchange (ETDEWEB)

    Hollnagel, Erik [Linkoeping Univ. (Sweden). Dept. of Computer and Information Science; Gauthereau, Vincent; Persson, Bodil [Linkoeping Univ. (Sweden). Dept. of Industrial Engineering

    2004-01-01

    This report describes the results from Phase III of a study on Operational Readiness Verification (ORV) that was carried out from December 2002 to November 2003. The work comprised a field study of ORV activities at a Swedish NPP during a planned productive outage, which allowed empirical work to be conducted in an appropriate environment with good accessibility to technical staff. One conclusion from Phase I of this project was the need to look more closely at the differences between three levels or types of tests that occur in ORV: object (component) test, system level test and (safety) function test, and to analyse the different steps of testing in order to understand the nontrivial relations between tests and safety. A second conclusion was the need to take a closer look at the organisation's ability to improvise in the sense of adjusting pre-defined plans to the actual conditions under which they are to be carried out. Phase II of the project found that although all three types of test occurred, they were rather used according to need rather than to a predefined arrangement or procedure. The complexity of ORV could be understood and described by using the concepts of Community of Practice, embedding, and Efficiency-Thoroughness Trade-Off. In addition, organisation and the different communities of practice improvise by adjusting pre-defined plans or work orders to the existing conditions. Such improvisations take place both on the levels of individual actions, on the level of communities of practice, and on the organisational level. The ability to improvise is practically a necessity for work to be carried out, but is also a potential risk. Phase III of the project studied how tasks are adapted relative to the different types of embedding and the degree of correspondence between nominal and actual ORV. It also looked further at the different Communities of Practice that are part of maintenance and ORV, focusing on the coordination and communication between

  17. The main trends of work on lifetime management of NPP

    International Nuclear Information System (INIS)

    Dragunov, Yu.; Kurakov, Yu.

    1995-01-01

    The major trends of works on NPP lifetime management are discussed: Co-ordinating and organizational activities; studying the degradation processes and residual life; elaboration and implementation of the measures to maintain/enhance safety level of NPPs operating within design life; development of control principals of advanced NPPs aging and safe life; creation and development of regulations in the field of NPP aging and life management. 7 figs

  18. International conference on the strengthening of nuclear safety in Eastern Europe. Keynote papers. Regulatory aspects of NPP safety, status of safety improvements, status of safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-06-01

    The Objective of the Conference was to assess the past decade of nuclear safety efforts in countries operating WWER and RBMK nuclear reactors and to address remaining safety issues which require further work. A particular focus of the Conference was on international co-operation and assistance and where such efforts should be focused in the future. All Eastern European countries that operate RBMK or WWER reactors participated in the Conference, and presented papers on three key areas of nuclear safety: Regulatory Aspects of Nuclear Power Plant Safety; Status of Safety Improvements; and Status of Safety Analysis Reports. In addition, representatives from 18 additional countries that provide financial and/or technical assistance and co-operation in the area of WWER and RBMK safety offered the most extensive commentary. Key international (IAEA, World Association of Nuclear Operators, the Nuclear Energy Agency, the G-24 NUSAC, the European Commission, and the EBRD) organizations that provide nuclear safety assistance for WWER and RBMK reactors also made presentations. There is no question that considerable progress on nuclear safety has been made in Eastern Europe. Special mention should be made of successful efforts to strengthen the independence and technical competence of the nuclear regulatory authorities. Efforts should now concentrate on improving the depth and scope of the technical abilities of the regulatory authorities. More attention by governments is needed to ensure that the regulatory authorities have the financial resources and enforcement authority to fully execute their missions. In respect to the operators of the nuclear power plants, they have demonstrated clear progress in operational safety improvements. Significant additional efforts are required to maintain and enhance an effective safety culture. Design safety improvement programmes are in place in all countries. Implementation of these programmes has varied and is particularly affected by

  19. International conference on the strengthening of nuclear safety in Eastern Europe. Keynote papers. Regulatory aspects of NPP safety, status of safety improvements, status of safety analysis report

    International Nuclear Information System (INIS)

    1999-06-01

    The Objective of the Conference was to assess the past decade of nuclear safety efforts in countries operating WWER and RBMK nuclear reactors and to address remaining safety issues which require further work. A particular focus of the Conference was on international co-operation and assistance and where such efforts should be focused in the future. All Eastern European countries that operate RBMK or WWER reactors participated in the Conference, and presented papers on three key areas of nuclear safety: Regulatory Aspects of Nuclear Power Plant Safety; Status of Safety Improvements; and Status of Safety Analysis Reports. In addition, representatives from 18 additional countries that provide financial and/or technical assistance and co-operation in the area of WWER and RBMK safety offered the most extensive commentary. Key international (IAEA, World Association of Nuclear Operators, the Nuclear Energy Agency, the G-24 NUSAC, the European Commission, and the EBRD) organizations that provide nuclear safety assistance for WWER and RBMK reactors also made presentations. There is no question that considerable progress on nuclear safety has been made in Eastern Europe. Special mention should be made of successful efforts to strengthen the independence and technical competence of the nuclear regulatory authorities. Efforts should now concentrate on improving the depth and scope of the technical abilities of the regulatory authorities. More attention by governments is needed to ensure that the regulatory authorities have the financial resources and enforcement authority to fully execute their missions. In respect to the operators of the nuclear power plants, they have demonstrated clear progress in operational safety improvements. Significant additional efforts are required to maintain and enhance an effective safety culture. Design safety improvement programmes are in place in all countries. Implementation of these programmes has varied and is particularly affected by

  20. How could intelligent safety transport systems enhance safety ?

    NARCIS (Netherlands)

    Wiethoff, M. Heijer, T. & Bekiaris, E.

    2017-01-01

    In Europe, many deaths and injured each years are the cost of today's road traffic. Therefore, it is wise to look for possible solutions for enhancing traffic safety. Some Advanced Driver Assistance Systems (ADAS) are expected to increase safety, but they may also evoke new safety hazards. Only

  1. Analysis of the network protection devices action connected with the Krsko NPP operation safety

    International Nuclear Information System (INIS)

    Omahen, P.; Struc, S.

    1995-01-01

    The example of failure that took place in Zagreb on 30 March, 1995, has been investigated in order to obtain appropriate answers to the unexpected electric power system (ESP) response and consecutive outage of the Krsko nuclear power plant (NPP). The analysis has been made of particular operating and stability conditions, related to the functioning of distance relays line protection. Using a consequent simulation model of the EPS (European interconnection) under specified fault condition, the research of EPS operation (relays action) and stability has been done. The appropriate simulation results have been compared to the available measured data which had been collected by the SOREL on-line data acquisition system. In the end, a practise proposal for measures to be taken with the target of achieving the foreseen and expected operation of the EPS equipment that effects operation of the Krsko NPP is given. (author)

  2. How to ensure the safety of extended operations: Practice and experience of Paks NPP

    International Nuclear Information System (INIS)

    Kovacs, J.

    2005-01-01

    The Paks Nuclear Power Plant strategy is to extend the operational lifetime of the plant and renew the operational license for 20 years over the designed and licensed lifetime. In the paper the preconditions of long-term operation are discussed and the basic findings and experience of the license renewal works are also presented. The further plans fo NPP Paks for ensuring safe operation in long-term are discussed. (author)

  3. Supporting the Ukrainian Supervisory authority when installing of an enhanced operational monitoring system for Saporoshje NPP. Final report

    International Nuclear Information System (INIS)

    Beyer, M.; Carl, H.; Reitz, T.

    2000-07-01

    The system to improve the operational monitoring of the unit 5 of the Zaporozh'ye NPP which was put into operation as a pilot project in the middle of the nineties has been extended to all six units and connected by means of modern communication hardware with the Information- and Crisis Centre (ICC) of the Ukrainian Supervisory Authority in Kiev. The technical means necessary for the completion of the local network in the Zaporozh'ye NPP were jointly specified, procured in Germany, tested in the necessary extent, consigned to the Ukraine and handed over free to the partner at the place of action. After an appropriate enlargement of the software for information processing, transfer and evaluation 159 safety related operational and 25 radiological measuring values of each of the six VVER-1000 units can be permanently recorded, online illustrated and evaluated on a computer in forms of diagrams, charts and graphs at the NPP site as well as transferred to the ICC. Thus, the same information quantity can be used for monitoring purposes in the ICC in Kiev. (orig.) [de

  4. Development of seismic safety reevaluation procedure considering the ageing of NPP facilities

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myoung Kue [Jeonju Univ., Cheonju (Korea, Republic of); Kim, J. M. [Cheonnam National Univ., Gwangju (Korea, Republic of); Kim, Y. S.; Cheong, S. H.; Kim, I. S.; Lee, M. G.; Kim, D. O. [Andong National Univ., Andong (Korea, Republic of); Lee, G. H. [Mokpo National Maritime Univ., Mokpo (Korea, Republic of)

    2003-03-15

    There are three of Nuclear Power Plants subject to the USI A-46 in Korea, including Kori No 1 and No 2 and Wolsung No 1. For the sake of resolution of the issue the possibility of adopting the GIP developed by the SQUG in USA is very high. In relation to the issue, this study addresses some technical improvements of the GIP including sloshing analysis based on multiple modes, seismic retrofit of cabinet for reduction of ICRS and modification of IRS depending on damping ratio. Dominant degradation factor and its affects NPP concrete elements are reviewed : chloride induced corrosion, carbonation of concrete elements, freezing and thawing of concrete elements, chemical and biological process, crack affect on concrete degradation. Various technical reports and papers about age-related degradation are reviewed for identification of degradation properties of NPP structures and components and degradation trend in NPP structures and components. This report summarizes numerical model for concrete degradation and development procedure of numerical models for concrete degradation. This report proposes the research necessity for performance evaluation of degraded concrete structure and selection of element for further study.

  5. Synergy in the areas of NPP nuclear safety and nuclear security

    International Nuclear Information System (INIS)

    Dybach, A.M.; Kuzmyak, I.Ya.; Kukhotskij, A.V.

    2013-01-01

    The paper considers the question of synergy between nuclear safety and nuclear security. Special attention is paid to identifying interface of the two areas of safety and definition of common principles for nuclear security and nuclear safety measures. The principles of defense in depth, safety culture and graded approach are analyzed in detail.Specific features characteristic of nuclear safety and security are outlined

  6. Generalization of Nuclear Safety and Course of Accident Events Research in the Ignalina NPP; Branduolines saugos ir avariju eigos Ignalinos AE tyrimu apibendrinimas

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, A; Uspuras, E [Lithuanian Energy Institute, Kaunas (Lithuania)

    2001-07-01

    The safety analysis shown that after implementation of SAR recommendations Ignalina NPP is adequately protected against accidents which required fast initiation of automatic protections. In case of accidents with long-term loss of core cooling additional operator actions are required. Accident management in case long-term core cooling are analyzed in this paper. (author)

  7. Results from probabilistic safety analysis level 2 for Kozloduy NPP units 3 and 4

    International Nuclear Information System (INIS)

    Velev, V.; Mancheva, K.

    2006-01-01

    The study considers full power operation analysis results and impact of strategies and modernisation implementations of Kozloduy NPP units 3 and 4. In this paper the following issues are discussed: the development of Severe Accident Management Guidelines and Symptom Based Emergency Operating Procedures, installation of brand new system as Emergency Filter Ventilation System, Passive Autocatalytic Hydrogen recombiners, etc. In order to respond to the objectives of the study, the analyses assets the impact of different initiating events on risk based on the frequencies of various plant damage states, the confinement failure and the radiological release frequencies and finally estimates the total risk as frequency of radiological releases

  8. Safety culture indicators for NPP: international trends and development status in Korea

    International Nuclear Information System (INIS)

    Choi, Y. S.; Ko, J. D.; Choi, K. S.; Jung, Y. H.

    2004-01-01

    Safety culture has been recognized as important to achieve high level of nuclear safety, as several recent events that have occurred in advanced countries were found to have important implications for safety culture. Under the recognition, implementation-focused and practical methods to foster safety culture have become necessary. Development of safety culture indicators for assessing the level of safety culture and identifying some deficiencies is being conducted. This paper examines the regulatory positions of major nuclear power countries on licensee's safety culture, introduces the development status of Korean Safety Culture Indicators and presents its future direction

  9. Influence of the CVCS Modelling on Results of the Loss of Offsite Power (LOOP) Safety Analysis for NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Bajs, T.; Debrecin, N.

    2006-01-01

    A Loss of Offsite Power (LOOP) transient scenario is based on a complete loss of non-emergency AC power that results in the loss of all power to the plant auxiliaries, i.e., the Reactor Coolant Pumps (RCPs), condensate pumps, etc. An actual LOOP event would cause a loss of all feedwater, a loss of forced Reactor Coolant System (RCS) flow and a reactor trip within less than 2 seconds as a result of either loss of power to the rod cluster assembly gripper coils or any RCS flow trips. For safety analysis purposes the LOOP event is conservatively modelled as a Loss of Normal Feedwater (LONF) transient with a subsequent loss of offsite power as a result of a reactor trip. The reactor trip followed by RCP trip are delayed until a low-low Steam Generator (SG) level signal is reached. This is a more conservative scenario than the LOOP event because the least amount of SG secondary side water mass available for heat removal and the increased amount of the stored energy in the primary circuit at the time of the loss of RCS flow result. The standard LOOP safety analysis is aimed to demonstrate the natural circulation capability of the RCS to remove residual and decay heat from the core aided by Auxiliary Feedwater in the secondary system. In addition to this goal the presented work is aimed to resolve the potential safety issue resulting from the influence of the Chemical and Volume Control System (CVCS) operation during LOOP event for NPP Krsko. The potential safety concern for the LOOP analysis is that the loss of instrument air system may occur thus leading to the CVCS charging and letdown flow imbalance. A net RCS inventory addition may result with water solid pressurizer condition. Water discharge through the pressurizer relief and safety valves could lead to overpressurization of the Pressurizer Relief Tank (PRT) and rupture of the PRT rupture disks. Additional concern is that pressurizer relief and safety valves may fail to properly reseat when exposed to water relief

  10. A usability review of a model checker VIS for the verification of NPP I and C system safety software

    International Nuclear Information System (INIS)

    Son, H. S.; Kwon, K. C.

    2002-01-01

    This paper discusses the usability of a model checker VIS in the verification of safety software of NPP I and C systems. The software development environment exemplified in this paper is for PLC and ESF-CCS which are being developed in KNICS project. In this environment, STATEMATE is used in requirement analysis and design phases. PLC is expected to be implemented using C language and an assembly language because it has many interfaces with hardware like CPU, I/O devices, communication devices. ESF-CCS is supposed to be developed in terms of PLC programming languages which are defined in IEC 61131-3 standard. In this case, VIS proved to be very useful through the review. We are also able to expect greater usability of VIS if we further develop the techniques for code abstraction and automatic translation from code to verilog, which is the input of VIS

  11. TVSA-T fuel assembly for 'Temelin' NPP. Main results of design and safety analyses. Trends of development

    International Nuclear Information System (INIS)

    Samojlov, O.B.; Kajdalov, V.B.; Falkov, A.A.; Bolnov, V.A.; Morozkin, O.N.; Molchanov, V.L.; Ugryumov, A.V.

    2010-01-01

    TVSA is a fuel assembly with rigid skeleton formed by 6 angle pieces and SG is successfully operated at 17 VVER-1000 power units of Kalinin NPP, as well as at Ukrainian and Bulgarian NPPs. Based on a contract for fuel supply to the Temelin NPP, the TVSA-T fuel assembly was developed, building on proven solutions confirmed by operation of TVSA modifications during 4-6 years and by the results of post-irradiation examination. The TVSA-T design includes combined spacer grids (SG+MG) and by fuel column elongation by 150 mm. A set of analyses and experiments was performed to validate the design, including thermal hydraulic tests, validation of critical heat flux correlation for TVSA-T, integrated mechanical, vibration and lifetime tests. A licence to use the fuel has been granted by the Czech State Office for Nuclear Safety. The TVSA-T core is currently in operation at the Temelin-1 reactor unit. The presentation is concluded as follows: TVSA-T fuel assembly for Temelin has been validated. The TVSA-T design is based on approved technical decisions and meets the current requirements for lifetime, operational maneuverability and safety. The results of post-irradiation examination of TVSA-T operated at the Kalinin-1 unit for 4 years confirm the assembly operability, skeleton stiffness, geometric stability and normal fuel rod cladding condition. The properties of the TVSA fuel with MG allow the core power to be increased up to 3300 MW to match the envisaged future VVER (MIR-1200) design, providing allowable fuel rod power FΔh =1.63 (to implement effective fuel cycles). (P.A.)

  12. EGP contribution to Mochovce completion, safety enhancement and operation

    International Nuclear Information System (INIS)

    Letko, A.; Matula, P.

    2000-01-01

    The Re-Evaluation Programme of Mochovce NPP was created in 1995 as an integral part of the completion of the Unit 1 and Unit 2. This program analyzed the general fulfillment of the principle of nuclear safety in the NPP Mochovce project. The analysis has required new corrections of the project, so that the project met the higher safety requirements when starting production. 87 safety measures represent the 'Program'. The basis for their creation were the international missions from 1992 to 1995 which defined. The final safety aim was represented by 'The Technical Specification of the Safety Measures' supported by The Nuclear Power Plant Research Institute and recommended by The Nuclear Regulatory Authority of the Slovak Republic. The technical specification served as a qualified base for the next steps in the pre-project, project and realization stages. (author)

  13. Regulatory approach to NPP ageing in Bulgaria

    International Nuclear Information System (INIS)

    Vassilev, D.

    2000-01-01

    In this contribution summary information of Kozloduy NPP units is presented. The nuclear legislation, regulatory approach for managing safety aspects on NPP ageing, short term programme, complex programme PRG'97 ant other aspects of ageing management are discussed

  14. Qualification of integrated tool environments (QUITE) for the development of computer-based safety systems in NPP

    International Nuclear Information System (INIS)

    Miedl, Horst

    2004-01-01

    In NPP I et C systems are back fitted meanwhile increasingly by computer-based systems (I et C platforms). The corresponding safety functions are implemented by software, and this software is developed, configured and administrated with the help of integrated tool environments (ITE). An ITE offers a set of services which are used to construct an I et C system and consist typically of software packages for project control and documentation, specification and design, automatic code generation and so on. Commercial ITE are not necessarily conceived and qualified (type-tested) for nuclear specific applications but are used - and will increasingly be used - for the implementation of nuclear safety related I et C systems. Therefor, it is necessary to qualify commercial ITE with respect to their influence on the quality of the target system for each I et C platform (dependent on the safety category of the target system). Examples for commercial ITEs are I et C platforms like SPINLINE 3, TELEPERM XP, Common Q, TRICON, etc. (Author)

  15. Suitable footwear for enhanced safety

    CERN Document Server

    2004-01-01

    Safety shoes are the theme of a new safety campaign. Always remember that accidents can happen - even to your feet! When entering hazardous areas such as underground halls, work sites, experiment assembly sites, workshops etc., sandals, ordinary shoes or similar light footwear should not be worn. Whatever the risks to which you may be exposed, always think safety and wear suitable footwear, i.e. safety shoes, which have non-slip soles and steel reinforcements to protect your feet from being crushed, fractured or pierced. Is it serious, Doctor? "Some traumas resulting from foot-related accidents - open fractures for instance - can be quite serious," explains CERN Works Doctor Véronique Fassnacht. "But the most common injuries are sprained ankles sustained during simple falls caused by differences in floor-levels (e.g. false floors). Fractures, bruising, surface wounds or deep wounds caused by objects falling onto the top of the foot are also quite common." ...

  16. Appraisal of the correspondence of the safety improvement measures proposed for the Mochovce NPP with respect to the ones outlined in the TEO for the Juragua NPP

    International Nuclear Information System (INIS)

    Pereira Hernandez, Guillermo

    1996-01-01

    The work accomplished consisted in determining the correspondence or relationship between the safety improvement measures of the outlined for both plants as well as a preliminary analysis from which were determined. The measures which a deeper analysis is recommended to evaluate theirs possible addition to the group it is recommended to deepen with a view to take into considerations aspects of interest that in them are outlined and which could serve to enhance the approach or the linkage of the measures outlined in the TEO

  17. Enhancement of nuclear safety culture

    International Nuclear Information System (INIS)

    Anderson, Stanley J.

    1996-01-01

    Throughout the 40-year history of the commercial nuclear power industry, improvements have continually been made in the design of nuclear power plants and the equipment in them. In one sense, we have reached an enviable point -- in most plants, equipment failures have become relatively rare. Yet events continue to occur. Regardless of how much the plants are improved, that equipment is operated by people -- highly motivated, well-trained people -- but people nonetheless. And people occasionally make mistakes. By setting the right climate and by setting high standards, good plant management can reduce the number of mistakes made ? and also reduce their potential consequences. Another way to say this is that the proper safety culture must be established and continually improved upon in our nuclear plants. Safety culture is defined by the International Atomic Energy Agency as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention that, as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance.' In short, we must make safety our top priority

  18. Enhancing Safety through Generic Competencies

    Directory of Open Access Journals (Sweden)

    S. Mockel

    2014-03-01

    Full Text Available This article provides insights into proactive safety management and mitigation. An analysis of accident reports reveals categories of supervening causes of accidents which can be directly linked to the concept of generic competencies (information management, communication and coordination, problem solving, and effect control. These findings strongly suggest adding the human element as another safety-constituting pillar to the concept of ship safety next to technology and regulation. We argue that the human element has unique abilities in dealing with critical and highly dynamic situations which can contribute to the system's recovery from non-routine or critical situations. By educating seafarers in generic competencies we claim to enable the people onboard to successfully deal with critical situations.

  19. Safety systems I/C equipment reliability analyses of the Kozloduy NPP units 3 and 4

    Energy Technology Data Exchange (ETDEWEB)

    Halev, G; Christov, N [Risk Engineering Ltd., Sofia (Bulgaria)

    1996-12-31

    The purpose of the analysis is to assess the safety systems I/C equipment reliability. The assessment includes: quantification of the safety systems unavailability due to component failures; definition of the minimal cut sets leading to the analysed safety systems failure; quantification of the I/C equipment importance measures of the dominant contribution components. The safety systems I/C equipment reliability has been analysed using PSAPACK (a code for probabilistic safety assessment). Fault trees for the following safety systems of the Kozloduy-3 and Kozloduy-4 reactors have been constructed: neutron flow control equipment, reactor protection system, main coolant pumps, pressurizer safety valves `Sempell`, steam dump systems, spray system, low pressure injection system, emergency feeding water system, essential service water system. THree separate reports have been issued containing the performed analyses and results. 1 ref.

  20. French regulatory approach to establishing the safety case for ageing NPP's

    International Nuclear Information System (INIS)

    Delage, M.

    1994-06-01

    The French regulatory procedures make provision for three main stages in the safety assessment of nuclear power plants. The first stage ends up with the construction licence and focuses on the assessment of the preliminary safety report. The second stage makes it possible to issue the fuel loading approval following evaluation of the provisional safety report. The third stage permits to declare the start of normal operation of the installation. The procedure, the tests and the assessment forming the overall strategy for safety regulations are described in detail. (R.P.)

  1. French regulatory approach to establishing the safety case for ageing NPP`s

    Energy Technology Data Exchange (ETDEWEB)

    Delage, M.

    1994-06-15

    The French regulatory procedures make provision for three main stages in the safety assessment of nuclear power plants. The first stage ends up with the construction licence and focuses on the assessment of the preliminary safety report. The second stage makes it possible to issue the fuel loading approval following evaluation of the provisional safety report. The third stage permits to declare the start of normal operation of the installation. The procedure, the tests and the assessment forming the overall strategy for safety regulations are described in detail. (R.P.).

  2. Monitoring of operational reliability of safety-related I and C subsystems at the Dukovany NPP

    International Nuclear Information System (INIS)

    Fuchs, P.; Sagl, P.; Zlamal, P.

    2007-01-01

    First, the situation existing in the data base in 1999, i.e. before the monitoring and the operational reliability monitoring concept were introduced, is highlighted. The technique of data processing is described with focus on the assessment of the relevancy of the records, component failure rate monitoring, estimation of basic statistical parameters, evaluation of the feasibility of component failure (or failure latency) detection, assessment of the mean time to repair, FMEA of the basic components (relays end measuring chains) to establish spurious signals and dangerous failure ratio. The reliability assessment of the system functions is based on structural reliability calculations (common cause failures not included). The outcomes from the operational reliability monitoring are presented in the form of a representative set of data, graphic charts and results of system function reliability assessment. Prospects for upgrading the I and C operational reliability monitoring system to the benefit of NPP Dukovany operating economy (life cycle costs evaluation, spare parts planning, RCM application) are outlined. (author)

  3. Building Newcomer Competence for NPP Safety Assessment through Learning by Doing: Development of Level 1 Probabilistic Safety Assessment for Research Reactors

    International Nuclear Information System (INIS)

    Kuzmina, Irina

    2014-01-01

    Final remarks: • COMPASS-M project is a very fruitful study. 1. State-of-the-art competence for PSA technique in Malaysia (applicable to nuclear installations, incl. RR and NPP). 2. PSA model and report for the operating research reactor in Malaysia. → Risk estimate of core damage and ranking contributors to the risk; → Basis for further safety improvement of RR as appropriate. 3. Input for IAEA’s publications on PSA for research reactors. • The results will be available to interested Member States (security considerations be addressed); → Completion in mid-2014, paper to be published in PSAM-12; ► Managerial support is instrumental for success of learning-by-doing projects

  4. A study of the international trend and comprehensive enhancement program on the Nuclear Power Plant safety

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Soon Hong; Cho, Nam Jin; Paek, Won Phil [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1990-12-15

    The objectives of this study are as follows : overview of the international trend related to the safety of Nuclear Power Plant(NPPs), study of the present status of NPP safety in Korea in aspects of design, construction and operation, suggestion of the comprehensive program to improve NPP safety in Korea. The results of this study can contribute to improve the safety of existing and future NPPs, and to establish the severe accident policy in Korea.

  5. A study of the international trend and comprehensive enhancement program on the Nuclear Power Plant safety

    International Nuclear Information System (INIS)

    Jang, Soon Hong; Cho, Nam Jin; Paek, Won Phil

    1990-12-01

    The objectives of this study are as follows : overview of the international trend related to the safety of Nuclear Power Plant(NPPs), study of the present status of NPP safety in Korea in aspects of design, construction and operation, suggestion of the comprehensive program to improve NPP safety in Korea. The results of this study can contribute to improve the safety of existing and future NPPs, and to establish the severe accident policy in Korea

  6. Development of an Owner Engineer's independent capability in NPP safety and licensing

    International Nuclear Information System (INIS)

    Auglaire, M.; Bayart, D.; D'Eer, A.; Polet, F.; Vanhoenacker, L.; Zhang, J.

    2002-01-01

    As Owner's Engineer to Electrabel, the Belgian utility which owns and operates the 7 NPPs in Belgium, Tractebel Energy Engineering has gained considerable experience in the field of ten-yearly safety overhauls of NPPs since 1983. It has developed a methodology leading to proposing corrective actions by means of a global and integrated approach in which safety improvement costs are optimized. Safety issues addressed during those projects encompass the writing of Probabilistic Safety Assessment studies, post-TMI recommendations implementation, the installation of autocatalytic recombiners, accident studies, protection against pressurized thermal shock, impact of flooding of internal or external origin, implementation of severe accident management guidelines, re-evaluation of the environment, verification of extreme climate conditions, updating of the Safety Analysis Reports, operation review. (author)

  7. Improving the safety of Ukrainian NPP to reach an internationally accepted level

    International Nuclear Information System (INIS)

    Bozhko, S.; Helske, J.; Janke, R.; Mayoral, C.

    2013-01-01

    This paper summarizes the safety status and the modernization progress of Ukrainian NPPs towards an internationally accepted level of safety. After a brief discussion of the concept of what is called an 'international accepted level' for new and operating NPPs, the status of Russian type WWER and in particular the Ukrainian NPPs is presented. Then, the performed investigations of the gaps between international accepted level and the original status of Ukrainian NPPs are presented. The safety objectives of the modernization programs, some examples of defence in depth improvements, and an overall view of the modernization programs of Ukrainian NPPs are produced. Then, few important safety improvements implemented at the oldest Ukrainian WWER-1000 South Ukraine-1 are given in more detail. Finally, a conclusion presents the current status on the way to fulfill the national safety targets and to reach an internationally accepted level for all the Ukrainian NPPs. The paper is followed by the slides of the presentation. (authors)

  8. Aims of failed fuel detection and substantiation of radiation safety at implementation of new kinds of nuclear fuel and fuel cycles on NPP with WWER

    International Nuclear Information System (INIS)

    Miglo, V.; Luzanova, L.

    2011-01-01

    Limiting of number of leaking fuel rods in a core during reactor operation in the analyses which are carried out for a substantiation of radiating safety for NPP with WWER as well as problems and possibilities of FFD at implementation of new kinds of fuel and fuel cycles are the main topics discussed in this paper. Available experience of designing of the NPP with WWER shows, that for ensuring of implementation of the RS criteria regarding limiting radioactive emissions from the NPP and doses of an irradiation of the population living near to NPP, it is required to regulate more rigidly number of failed fuel rods in comparison with requirements of Rules of nuclear safety NP-082-07. The reason of it is necessity to consider a technical condition of all safety barriers on a path of radioactive FP extension in a complex, first and foremost of uncontrolled leakage of the primary coolant to the NPP premises and efficiency of filters of ventilating systems, and also spike-effect on activity of isotopes of iodine after a power unit shutdown for fuel reloading and openings of a cover of a reactor. Depending on the project of NPP, parameters of fuel loading, a place of placing of the NPP and other factors the limit level of activity of isotopes of iodine in the primary coolant will be reached at various number of leaking fuel rods which can be unequal for various power units and the NPP with WWER, constructed on one design. The quantity of leaking fuel rods at which the design limit on FP-activity in the primary coolant of operating reactor is reached, can be essential below an operational limit on number of failed fuel rods established by Rules of nuclear safety. However the reached quality of fabrication of the WWER fuel rods providing their high reliability (the probability of fuel rod failure in the course of one operation year is not higher than 10 -5 ) as well as due to the levels of the WWER fuel rod depressurization actually attainable in the normal conditions of

  9. A new design concept for offshore nuclear power plants with enhanced safety features

    International Nuclear Information System (INIS)

    Lee, Kihwan; Lee, Kang-Heon; Lee, Jeong Ik; Jeong, Yong Hoon; Lee, Phill-Seung

    2013-01-01

    Highlights: ► A new design concept for offshore nuclear power plants is proposed. ► The total general arrangement for the concept is suggested. ► A new emergency passive containment cooling system (EPCCS) is proposed. ► A new emergency passive reactor-vessel cooling system (EPRVCS) is proposed. ► Safety features against earthquakes, tsunamis, and storms are discussed. - Abstract: In this paper, we present a new concept for offshore nuclear power plants (ONPP) with enhanced safety features. The design concept of a nuclear power plant (NPP) mounted on gravity-based structures (GBSs), which are widely used offshore structures, is proposed first. To demonstrate the feasibility of the concept, a large-scale land-based nuclear power plant model APR1400, which is the most recent NPP model in the Republic of Korea, is mounted on a GBS while minimizing modification to the original features of APR1400. A new total general arrangement (GA) and basic design principles are proposed and can be directly applied to any existing land based large scale NPPs. The proposed concept will enhance the safety of a NPP due to several aspects. A new emergency passive containment cooling system (EPCCS) and emergency passive reactor-vessel cooling system (EPRVCS) are proposed; their features of using seawater as coolant and safety features against earthquakes, Tsunamis, storms, and marine collisions are also described. We believe that the proposed offshore nuclear power plant is more robust than conventional land-based nuclear power plants and it has strong potential to provide great opportunities in nuclear power industries by decoupling the site of construction and that of installation.

  10. RAF/9/049: Enhancing and Sustaining the National Regulatory Bodies for safety

    International Nuclear Information System (INIS)

    Keter, C.J.

    2017-01-01

    The main objective of this project is to enhance regulatory infrastructure, sustainability and cooperation among national regulatory bodies. This will support strengthening of the existing regulatory framework and capacity building in the region. Self-Assessment using the Self-Assessment Regulatory Infrastructure for Safety (SARIS) was completed on 26th May 2016. Changes made to the legislation is ongoing. The Nuclear Regulatory Bill 2017 is at an advanced stage and about to be tabled to Cabinet. The project objectives shall be addressed under a new project, RAF/9/058 – Improving the Regulatory Framework for the Control of Radiation Sources in Member States. Two major tasks for Kenya to focus include Review of regulations on waste safety, radiation sources and on safety of NPP and advising on drafting of radiation safety guides

  11. Ageing study of the engineered safety features actuation system of the Loviisa NPP

    International Nuclear Information System (INIS)

    Simola, K.; Maskuniitty, M.

    1995-06-01

    An ageing study of the engineered safety features actuation system of the Loviisa nuclear power plant has been performed. The operating experience, including failure and maintenance histories of analog measuring devices, logics for safety signal formation and individual control electronics of pumps and valves, has been collected and analysed. The safety importance of system components has been studied with a fault tree analysis of a selected safety function. Based on the results of the analysis of operating experiences and the fault tree analysis, some components were selected for deeper analyses. According to the operating experience, the amount of failures in the Loviisa plant safety system has been low and no increasing trend in the failure history can yet be observed. Only a few failures had prohibited the propagation of the safety signal, mostly the failures have caused a false alarm. The failures reported have concerned mainly limit signal units, transmitters, and priority units. According to the fault tree analysis of one safety function, the most important components of this subsystem are individual control units and pulse/DC converters. Failure modes and effect analyses were performed for priority and individual control unit, limit signal unit and comparator and pulse/DC converter in order to identify the critical failure modes of these devices. (orig.) (15 refs., 26 figs., 9 tabs.)

  12. Some aspects of RF radiation safety guidelines on urgent protective measures in case of radiation emergency at NPP

    International Nuclear Information System (INIS)

    Bulgakov, V.G.; Klepikova, N.V.; Shershakov, V.M.; Ivanov, E.A.

    2003-01-01

    VR or RBMK reactors using the forces of NPP before radiation safety experts become involved. This method was developed with the purpose to 1.) Improve radically fast response in terms of preparation of recommendations an intervention measures in the first hours into an accident. 2.) Reduce stress for the NPP administration and liberate them for efficient management of an accident The task of preparing recommendations an emergency measures in the first hours of an accident by NPP forces can be efficiently executed based on preliminary study of a set of initial data, which am key for radiation consequences of an accident and developing a set of recommendations for each set. The presentation describes a method for selection of a set of initial data and identifying corresponding areas in which emergency measures would be reasonable to implement. The considered key factors influencing the radiation situation in the early phase of an accident at NPP include: the activity of 131 I release to the atmosphere (contribution of other iodine radioisotopes can be allowed for by using a correction factor), physical and chemical forms of iodine in the release, a release duration, time of the day during which the release is dispersed in the atmosphere, the population age group, the atmospheric boundary layer stability category, wind speed and the underlying surface roughness length. In selecting a set of initial data preference was given to standard information about meteorological conditions of pollutant dispersion in the atmosphere available to meteorological stations and dosimetry services of NPPs. Analysis was performed for joint realization of atmospheric stability category and classification of wind speed at vane's height. Results of analysis are presented. tab. 1 (author)

  13. Seismic safety review mission for the follow-up of the seismic upgrading of Kozloduy NPP (Units 1-4). Sofia, Bulgaria, 16-20 November 1992

    International Nuclear Information System (INIS)

    David, M.; Shibata, H.; Stevenson, J.D.; Godoy, A.; Gurpinar, A.

    1992-11-01

    A Seismic Safety Review Mission for the follow-up of the design and implementation of the seismic upgrading of Kozloduy NPP was performed in Sofia from 16-20 November 1992. This mission continued the second task of the follow-up activities of the design and implementation of the seismic upgrading (Phases 1 and 2), which is being carried out in Units 1 and 2 of the NPP. Thus the objectives of the mission was to assist the Bulgarian authorities in the technical evaluation of the design tasks defined for Phases 1 and 2 item HB of WANO 6 Month Programme, as follows: anchorage upgrades of low seismic capacity components; list of seismic safety related systems and components; detailed walkdown to assess seismic capacity of components and define priorities for the upgrading; determination of seismic structural capacity of pump house, diesel generator building and turbine building and design of required upgrades; liquefaction potential evaluation. Tabs

  14. Safety enhancement efforts after Fukushima accident in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, U.C., E-mail: uclee@nssc.go.kr [Nuclear Safety & Security Commission, Seoul (Korea, Republic of)

    2014-07-01

    On March 11 of 2011, a massive earthquake and powerful tsunami hit the north-eastern region of Japan and the Fukushima Daiichi Nuclear Power Plant was massively damaged. Korea which is located closest to Japan was not directly affected, however, its people were shocked. They were concerned over the possibility of being exposed to radiation as well as for the safety of domestic nuclear power plants. The Korean government recognized the need to take prompt and immediate actions to alleviate these concerns. The Korean government immediately implemented special safety inspection and derived 50 long and short-term improvement action items to ensure safety of NPPs under extreme hazard conditions. At present, stage 3 of implementation strategy is being implemented, with completion of 22 items including ASTS (Automatic Seismic Trip System) as well as revision of 14 items including suitability review of action measures taken for investigation and research of maximum earthquake at NPP sites. The IAEA Integrated Regulatory Review Service (IRRS) mission was conducted for two weeks during July 10 and 22, 2011, which happened to be the very first review mission to be carried out since the Fukushima disaster. A module on the policy issues related to the actions taken after the Fukushima accident was newly added to the mission. The mission highlighted positive aspects of Korea's safety regulatory program, praising its technical competence and effectiveness. Additionally it concluded that Korea has been responding to the accident in a timely and an effective manner. The follow-up review mission is scheduled in December of this year, expanding the scope to include radiation safety as well. The most noteworthy change in Korea since the Fukushima accident is independence of the regulatory body. Not only to enhance effectiveness and independence of the regulatory body but to secure nuclear safety, the Nuclear Safety and Security Commission (NSSC) was established on October 26

  15. Romania - NPP PLiM Between Regulatory Requirement / Oversight and Operator Safety / Financial Interest

    International Nuclear Information System (INIS)

    Goicea, Lucian

    2012-01-01

    Cernavoda Unit 1 PLiM started in the first third of its design life, to develop as regulatory requirements of the components of standards and programmes and to benefit by earlier implementation of the measures for achieving maximum operating life. CNCAN regulatory present approach on the utility PLiM combines the regulatory requirements on management system, ageing management provisions of periodic safety review, detailed technical requirements of ageing programmes and different techniques focusing only on safety issues. (author)

  16. Impact of changed fuel performances on safety barrier effectiveness at normal operation of NPP with WWER

    International Nuclear Information System (INIS)

    Zhurbenko, A. V.; Semchenkov, Y. M.; Slavyagin, P.D.

    2007-01-01

    The paper presents the analysis of adopted safety barriers against propagation of fission product released from WWER core of active power plants. Relationship between system and equipment performances and safety barriers is demonstrated. The fundamental principles of methodological approach to the operational limit determination based on the assessment of iodine-131 specific activity in the primary circuit are discussed. Problems of substantiating the operational limit for primary coolant activity are analyzed for conditions of growing burnup (Authors)

  17. Mechanical Property Characteristics of Butt-Fusion Joint of High Density Polyethylene Pipe for NPP Safety Class Application

    International Nuclear Information System (INIS)

    Oh, Youngjin; Kim, Kyoungsu; Lee, Seunggun; Park, Heungbae; Yu, Jeongho; Kim, Jongsung; Kim, Jeonghyun; Jang, Changheui; Choi, Sunwoong

    2013-01-01

    Several NPPs in United States replaced parts of sea water or raw water system pipes to HDPE (high density polyethylene) pipes, which have outstanding resistance for oxidation and seismic loading. ASME B and PV code committee developed Code Case N-755, which describes rules for the construction of Safety Class 3 polyethylene pressure piping components. Several NPP's in US proposed relief requests in order to apply Code Case N-755. Although US NRC permitted using Code Case N-755 and HDPE materials for Class 3 buried piping, their permission was limited to only 10 years because of several concerns for material performance of HDPE. US NRC's major concerns are about material properties and the quality of fusion zone of HDPE. In this study, material property tests for HDPE fusion zone are conducted with varying standard fusion procedures. Mechanical property tests for fused material for HDPE pipes were conducted. Fused material shows lower toughness than base material and fused material of lower fusion pressure shows higher toughness than that of higher fusion pressure

  18. Seismic assessment of safety-related structures: laboratory testing of the pressure relief duct frame at pickering NPP

    International Nuclear Information System (INIS)

    Ghobarah, A.; Biddah, A.; Pilette, C.

    1995-01-01

    The pressure relief duct (PRD) is a Special safety System in the CANDU-PHW multi-unit nuclear power plants (NPP). It is designed to contain and direct the outflow from the reactor building to the pressure suppression and containing systems in the vacuum building. The PRD is a large elevated reinforced concrete box structure of internal width of 6.1 m, height of 7.6 m, and wall thickness of 0.6 m. The PRD is 662 m long and is supported every 22 m by concrete frames of height of 21 m. Typical frame members are 1.8 m in depth and width. A representative elevation of the frame is presented. The section of the PRD under investigation was designed and constructed before the current seismic design codes were in effect. An assessment of the PRD structure when subjected to various levels of ground motion has shown that the frame has a limited seismic withstand capacity. Its seismic performance is dependent on the ductility of the beams and on the ability of the beam-column joint to transfer bending moments and shear forces. The objectives of this study are to provide the data to validate the frame analysis results through laboratory testing of a scaled specimen of the beam-column joint, and to compare the observed response with the response of a beam-column joint when the shear reinforcement is detailed according to current seismic design codes. (author). 3 refs., 10 figs

  19. End of mission report on seismic safety review mission for Belene NPP site

    International Nuclear Information System (INIS)

    Gurpinar, A.; Mohammadioun, B.; Schneider, H.; Serva, L.

    1995-01-01

    Upon the invitation of the Bulgarian government through the Committee for the Peaceful Uses of Atomic Energy and within the framework of the implementation of the Technical Cooperation project BUL/9/012 related to site and seismic of NPPs, a mission visited Sofia 3 - 7 July 1995. The mission constituted a follow-up of the interim review of subjects related to tectonic stability and seismic hazard characterization of the site which was performed in September 1993. The main objective of the mission was the final review of the subjects already reviewed in September 1993 as well as issues related to geotechnical engineering and foundation safety. The main terms of reference of the present mission was to verify the implementation of the recommendations of the Site Safety Review Mission of June 1990. This document gives findings on geology-tectonics, seismology and foundation safety. In the end conclusions and recommendations of the mission are presented

  20. Report of a consultants meeting on backfittings and safety enhancement measures in NPPs with WWER 440/213 reactors. Extrabudgetary programme on the safety of WWER NPPS

    International Nuclear Information System (INIS)

    1994-01-01

    The purpose of this Consultants' Meeting held by the IAEA in Vienna from 11-15 April 1994 within the framework of the Extrabudgetary Programme on WWER Safety was to review and analyze safety issues revealed during operation and through analyses of NPPs with WWER 440/213 reactors. The initial list of safety issues based on the available reports from various studies had been prepared by the IAEA secretariat before the meeting, together with indications of safety enhancement measures proposed in various NPP units. During the meeting, the underlying safety concerns and actual technical status of the plants were discussed and the ranking of the safety issues was considered. 58 refs, 1 tab

  1. Enhancement and validation of the NPP Mühleberg MCNP activation simulations for Swiss decommissioning planning

    International Nuclear Information System (INIS)

    Bykov, V.

    2014-08-01

    The Swiss National Cooperative for the Disposal of Radioactive Waste (NAGRA) regularly performs analysis of cost estimates associated with the NPP decommissioning. For this purpose, NAGRA has over the past ten years developed a NPP activation analysis methodology based on MCNP models of Swiss NPPs. The validation of these models is accomplished using measurements from oil activation campaigns, in which foil samples are activated at key locations inside the NPP for the duration of one cycle. The measurement campaigns have already been carried out at the Gösgen PWR (KKG) and the Mühleberg BWR (KKM). The first validation has already been successfully conducted for the KKG MCNP model. This thesis describes the efforts to validate the KKM MCNP model. This process included modifications, such as modeling of steam separators individually and improving the definition of jet pumps. Furthermore, the core definition was completely redefined, going from a 6-cell cylindrical model to a 940-cell model, shaped like the actual KKM core, which more accurately represented the void distribution. In order to benchmark the new model, the locations of samples during the two KKM foil activation campaigns were implemented into the model using the GSAM code. The interface between the MCNP model and GSAM was improved by creating a new energy group structure, optimized specifically for the activation of the three foil materials. Their activation was stimulated the state of the art hybrid VR code ADVANTG. The calculated results were then compared against the measured values for each foil material separately. The numerous improvements introduced in the 2014 model led to good agreement in many areas. The agreement is within the factor of two on the inner side of the bioshield, at the core height and above, and factor of three above the bioshield. Furthermore, distinct suggestion for improving the agreement in other areas was presented. This includes modeling of pipes extending from the RPV

  2. Presentation of a method for the sequential analysis of incidents - NPP safety

    Energy Technology Data Exchange (ETDEWEB)

    Delage, M; Giroux, C; Quentin, P

    1989-04-01

    This paper presents a method which is designed to assist in the analysis of safety and based on the graphic representation of the occurrence of incidents significant for safety in 900-MWe PWR units. The graphs obtained are linked together to produce a general tree of events. With this tool, and on the basis of operating experience, it is then possible to imagine complex incident scenarios, to evaluate the potential consequences of a particular incident, or to seed out the causes which could lead to a given event. Interactions between systems or common mode faults can also be evidenced with this method.

  3. Application of the Combined Cycle LWR-Gas Turbine to PWR for NPP Life Extension Safety Upgrade and Improving Economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu. N.

    2006-01-01

    Currently, some of the most important problem for the nuclear industry are life extension, advance competitiveness and safety of aging LWR NPPs. Based on results of studies performed in the USA (Battelle Memorial Institute) and in Russia (NIKIET), a new power technology, using a combined cycle gas-turbine facility CCGT - LWR, so called TD-Cycle, can significantly help in resolution of some problems of nuclear power industry. The nuclear steam and gas topping cycle is used for re-powering a light water pressurized reactor of PWR or VVER type. An existing NPP is topped with a gas turbine facility with a heat recovery steam generator (HRSG) generating steam from waste heat. The superheated steam of high pressure (P=90-165 bar, T=500-550 C) generated in the HRSG, is expanded in a high pressure (HP) turbine for producing electricity. The HP turbine can work on one shaft with the the gas turbine or at one shaft with intermediate (IP) or low (LP) pressure parts of the main nuclear steam turbine, or with a separate electric generator. The exhausted steam from the HP turbine is injected into the steam mixer where it is mixed with the saturated steam from the NPP steam generator (SG). The mixer is intended to superheat the main nuclear steam and should be characterized by minimum losses during mixing superheated and saturated steam. Steam from the mixer superheated by 20-60 C directs to the existing IP turbine, and then, through a separator-reheater flows into the LP turbine. Feed water re-heaters of LP and HP are actually unchanged in this case. Feed water extraction to the HRSG is supplied after one of LP water heaters. This proposal is intended to re-power existing LWR NPPs. To minimize cost, the IP and LP turbines and electric generator would remain the same. The reactor thermal power and fast neutron flux to the reactor vessel would decrease by 30-50 percent of nominal values. The external peripheral row of fuel elements can be replaced with metal absorber rods to

  4. Nuclear Safety at the Ignalina NPP. Achievements and challenges. Proceedings of the international conference

    International Nuclear Information System (INIS)

    1999-01-01

    The International Conference was held on 8-9 April, 1999 in Vilnius. The aim of this conference - to offers specialists an excellent chance to get acquainted with Lithuania's experience and the work it has done in the area of nuclear safety. Also this conference will help to depoliticize the discussion of nuclear power plants

  5. Safety assessment of multi-unit NPP sites subject to external events

    International Nuclear Information System (INIS)

    Samaddar, Sujit; Hibino, Kenta; Coman, Ovidiu

    2014-01-01

    This paper presents a framework for conducting a probabilistic safety assessment of multi-unit sites against external events. The treatment of multiple hazard on a unit, interaction between units, implementation of severe accident measures, human reliability, environmental conditions, metric of risk for both reactor and non-reactor sources, integration of risk and responses and many such important factors need to be addressed within the context of this framework. The framework facilitates the establishment of a comprehensive methodology that can be applied internationally to the peer review of safety assessment of multi-unit sites under the impact of multiple external hazards. In summary, it can be said that the site safety assessment for a multi-unit site will be quite complex and need to start with individual unit risk assessments, these need to be combined considering the interactions between units and their responses, and the fragilities of the installations established considering the combined demands from all interactions. Using newly established risk metric the risk can then be integrated for the overall site. Fig. 2 shows schematically such a proposal. Much work has to done and the IAEA has established a working group that is systematically establishing the structure and process to incorporate the many issues that are a part of a multi-unit site safety assessment. (authors)

  6. Structural analysis strategies of the pressurized relief and safety valves discharge piping of NPP Angra 1

    International Nuclear Information System (INIS)

    Lima, Maria Ines Prates de; Kuramoto, Edson; Suanno, Rodolfo

    2002-01-01

    The pressurizer relief and safety valve system provides the reactor coolant system overpressure protection and, therefore, it is fundamental for the security of a nuclear plant. This paper discusses the safety valve loop seal strategies adopted by others nuclear power plants over the world in order to attend the recommendations of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations). The technical option adopted for Angra 1 consists in making specific modifications on the original piping and support configuration of the pressurizer relief and safety valve system. These modifications were proposed in order to reduce the high stress levels induced by the thermal-hydrodynamic loads caused by the discharge of the sub-cooled water during the opening of the relief or the safety valves. Several thermal-hydraulic models were tested to assess the influence of the seal water heating and the simultaneous opening of the valves in order to minimize the thermal hydrodynamic loads effects. The piping structural analysis was performed, using the computer program system KWUROHR, to satisfy the requirements of the appropriate equations of the code ASME Section III, Subsections NB3650 and NC3650. (author)

  7. Initial data on protective containment for safety analysis of NPP with WWER-1000 Preparation

    Directory of Open Access Journals (Sweden)

    Gubeladze Oleg

    2017-01-01

    Full Text Available In the presented article the nuclear safety issues with VVER-1000 reactor are considered. The study element is the protective containment (PC, the most important function of which is localization and retention of radioactive substances within the accident localization zone. The example of possible unregulated destructive forcing (UDF on the PC for the construction and installation work period is given.

  8. European type NPP electric power and vent systems. For safety improvement and proposal of international center

    International Nuclear Information System (INIS)

    Sugiyama, Kenichiro

    2011-01-01

    For prevention of reactor accidents of nuclear power plants, multiplicity and redundancy of emergency power would be most important. At station blackout accident, European type manually operated vent operation could minimize release amount of radioactive materials and keep safety of neighboring residents. After Fukushima Daiichi accident, nuclear power plants could not restart operation even after completion of periodical inspection. This article introduced European type emergency power and vent systems in Swiss, Sweden and Germany with state of nuclear power phaseout for reference at considering to upgrade safety and accident mitigation measures for better understanding of the public. In addition, it would be important to recover trust of nuclear technology to continue to disseminate latest information on new knowledge of accident site and decontamination technologies to domestic and overseas people. As its implementation, establishment of Fukushima international center was proposed. (T. Tanaka)

  9. Overview of IRSN R and D on NPP safety, with focus on severe accident

    International Nuclear Information System (INIS)

    Van Dorsselaere, Jean-Pierre

    2015-01-01

    IRSN contributes to the continuous improvement of safety level of Gen.II and III reactors, with the aim to approach for Gen.II the target safety level of Gen.III. This needs to build the necessary knowledge to appreciate margins for safety important systems, structures and components in the frame of plant operation life extension beyond 40 years. Research is a major IRSN mission that is tightly linked to expertise needs: it involves 40% of overall budget and, out of radiation protection and safety of waste disposal, around 280 scientists. IRSN has acquired a huge experience in the last 30 years on severe accidents, both on experimental and theoretical aspects, in particular through management of large international research programmes like the Phébus. FP integral experiments in the last 20 years and the coordination of the SARNET network of excellence that continues now in the frame of the NUGENIA European association. Besides, IRSN is developing, in collaboration with GRS (Germany), the integral system code ASTEC that is considered now as the European reference code due to the continuous capitalization of all the international knowledge. The presentation summarizes the ongoing IRSN research on the different phenomena involved in severe accidents, with more focus in the last years on mitigation devices or measures, i.e. for in-vessel and ex-vessel corium coolability, hydrogen explosion risk and source term. IRSN leads several international projects in Euratom frame (such as CESAM on ASTEC, PASSAM on source term mitigation, and IVMR on in-vessel corium retention) or OECD/NEA/CSNI (such as STEM). Moreover, several national projects on the above issues are ongoing with the French actors in this domain. Collaboration between IRSN and India is very active and efficient on ASTEC code with BARC and AERB, in particular through PHWR model development and assessment, and could be extended in the future to other issues either on severe accidents or on other Topics. (author)

  10. Risk informed approach and its application in Daya Bay NPP operation safety management

    International Nuclear Information System (INIS)

    He Yu; Zhang Jinlong; Bao Yukun

    2004-01-01

    The paper presents a systematic risk assessment approach based on probabilistic theory, and discusses its significance and application process in safety management. Risk informed approach that uses deterministic engineering principles and probabilistic methods is the appropriate approach to decision making at nuclear power plants. The paper also studies an actual case taken place at Daya Bay Nuclear Power Station using PSA approach to equipment maintenance. (authors)

  11. Dukovany NPP maintenance management

    International Nuclear Information System (INIS)

    Siegel, F.

    2005-01-01

    Maintenance planning and management for the technological equipment of the Dukovany NPP are described. A Maintenance Control and Evaluation System has been developed and is in use to help manage the complex maintenance issue. Practical examples of outputs of the System, with a comprehensive use of the cost, reliability and safety related data, are presented. (author)

  12. Seismic safety review mission to assist in the evaluation of the design of seismic upgrading for Kozloduy NPP. Sofia, Bulgaria, 19-23 October 1992

    International Nuclear Information System (INIS)

    Ma, D.; Prato, C.; Godoy, A.

    1992-10-01

    A seismic Safety Review Mission to assist in the evaluation of the design of seismic upgrading for Kozloduy NPP was performed in Sofia from 19-23 October 1992. The objectives of the mission were to assist the Bulgarian authorities in: the evaluation of the floor response spectra of the main buildings of units 1-4 at Kozloduy NPP, calculated for the new defined seismic parameters at site (Review Level Earthquake - RLE); the evaluation of the remedial and strengthening measures proposed for the seismic upgrading of the pump house and diesel generator buildings to the new defined RLE. This mission completed the scope of previous IAEA mission - BUL/9/012-18b - (see Report 3262) performed from 3-7 August 1992, with regard to tasks which were not evaluated at that time because they had not been finished. 2 tabs

  13. Validation of programmable industrial automation systems for safety critical applications in NPP's dynamic testing

    International Nuclear Information System (INIS)

    Haapanen, P.; Korhonen, J.

    1995-01-01

    The safety assessment of programmable automation systems can not totally be based on conventional probabilistic methods because of the difficulties in quantification of the reliability of the software as well as the hardware. Additional means shall therefore be used to gain more confidence on the system dependability. One central confidence building measure is the independent dynamic testing of the completed system. An automated test harness is needed to run the required large amount of test cases in a restricted time span. The prototype dynamic testing harness for programmable digital systems developed at the Technical Research Centre of Finland (VTT) is described in the presentation. (12 refs., 2 figs., 2 tabs.)

  14. Validation of programmable industrial automation systems for safety critical applications in NPP's; dynamic testing

    International Nuclear Information System (INIS)

    Haapanen, P.; Korhonen, J.

    1995-01-01

    The safety assessment of programmable automation systems cannot be totally be based on conventional probabilistic methods because of the difficulties in quantification of the reliability of the software as well as the hardware. Additional means shall therefore be used to gain more confidence on the system dependability. One central confidence building measure is the independent dynamic testing of the completed system. An automated test harness is needed to run the required large amount of test cases in a restricted time span. This paper describes a prototype dynamic testing harness for programmable digital systems developed at VTT. (author). 12 refs, 2 figs, 2 tabs

  15. Safety Analysis in Design and Assessment of the Physical Protection of the OKG NPP

    Energy Technology Data Exchange (ETDEWEB)

    Lindahl, P., E-mail: par.lindahl@okg.eon.se [OKG Aktiebolag, Oskarshamn (Sweden)

    2014-10-15

    OKG AB operates a three unit nuclear power plant in the southern parts of Sweden. As a result of recent development of the legislation regarding physical protection of nuclear facilities, OKG has upgraded the protection against antagonistic actions. The new legislation includes requirements both on specific protective measures and on the performance of the physical protection as a whole. In short, the performance related requirements state that sufficient measures shall be implemented to protect against antagonistic actions, as defined by the regulator in the “Design Basis Threat” (DBT). Historically, physical protection and nuclear safety has been managed much as separate issues with different, sometimes contradicting, objectives. Now, insights from the work with the security upgrade have emphasized that physical protection needs to be regarded as an important part of the Defence-In-Depth (DiD) against nuclear accidents. Specifically, OKG has developed new DBT-based analysis methods, which may be characterized as probabilistically informed deterministic analysis, conformed to a format similar to the one used for conventional internal events analysis. The result is a powerful tool for design and assessment of the performance of the protection against antagonistic actions, using a nuclear safety perspective. (author)

  16. Assessment of the Zaporizhya NPP unit 1 reactor pressure vessel safety

    International Nuclear Information System (INIS)

    Podkopaev, V.; Popov, V.; Zaritsky, N.

    1997-01-01

    This emergency situation had occurred at the ZNPP unit 1 while its being under ''hot shutdown'' in natural coolant circulation mode. The main difference between emergency situation and mode with improper setting of PPPD described in the ''Technical Safety Substantiation (TSS) is that this mode is being considered in the TSS under rated power of reactor with main circulation pumps (MCP) under operation. This difference is a substantial one. For this reason a necessity appeared to asses an integrity of referred reactor pressure vessel (RPV) under given emergency situation to judge whether results obtained meet the ND requirements (safety assessment). Under operation such RPV elements are being mostly affected as upper cooling, lower cowling, weld No. 3 weld No. 4 situated in front of core. These elements materials ageing process is the most intense one. Thus, this work was aimed at investigation of structure material behavior and RPV integrity assessment under thermal shock conditions while PPPD improper setting. At that time the most attention was drawn to above mentioned upper and lower cowlings along with welds No. 3 and 4. 5 refs, figs, 10 tabs

  17. Assessment of the Zaporizhya NPP unit 1 reactor pressure vessel safety

    Energy Technology Data Exchange (ETDEWEB)

    Podkopaev, V; Popov, V; Zaritsky, N [State Scientific and Technical Centre on Nuclear and Radiation Safety (SSTC NRS), Kiev (Ukraine)

    1997-09-01

    This emergency situation had occurred at the ZNPP unit 1 while its being under ``hot shutdown`` in natural coolant circulation mode. The main difference between emergency situation and mode with improper setting of PPPD described in the ``Technical Safety Substantiation (TSS) is that this mode is being considered in the TSS under rated power of reactor with main circulation pumps (MCP) under operation. This difference is a substantial one. For this reason a necessity appeared to asses an integrity of referred reactor pressure vessel (RPV) under given emergency situation to judge whether results obtained meet the ND requirements (safety assessment). Under operation such RPV elements are being mostly affected as upper cooling, lower cowling, weld No. 3 weld No. 4 situated in front of core. These elements materials ageing process is the most intense one. Thus, this work was aimed at investigation of structure material behavior and RPV integrity assessment under thermal shock conditions while PPPD improper setting. At that time the most attention was drawn to above mentioned upper and lower cowlings along with welds No. 3 and 4. 5 refs, figs, 10 tabs.

  18. 78 FR 53790 - Public Forum-Safety Culture: Enhancing Transportation Safety

    Science.gov (United States)

    2013-08-30

    ... NATIONAL TRANSPORTATION SAFETY BOARD Public Forum--Safety Culture: Enhancing Transportation Safety On Tuesday and Wednesday, September 10-11, 2013, the National Transportation Safety Board (NTSB) will convene a forum titled, ``Safety Culture: Enhancing Transportation Safety.'' The forum will begin at 9:00...

  19. TELEPERM XS: I and C systems for safety application in NPP's - features, developments, references and feedback

    International Nuclear Information System (INIS)

    Prehler, Heinz Josef

    2007-01-01

    In the field of digital I and C AREVA NP is focused on concepts that on the one hand make allowance for development cycles getting shorter in the technology competition and on the other hand assure a long-term system support with the ability to deliver spare parts in the long run. The system platform TELEPERM XS, which was developed especially for safety I and C application of nuclear power plants, meets requirements effectively and thus provides a great benefit for the customer. The typical applications of TELEPERM XS are in the field of Reactor Protection and ESFAS functions (Engineered Safety Features Actuation System). High demands are defined for system reliability and availability, as well as for failure prevention and tolerance. The requirements of corresponding international codes and standards of nuclear installations are also implemented in the development and engineering processes of TELEPERM XS. The system platform is integrated into a sustainable program for service life management of electronic systems and equipment. Its ongoing future-oriented development ensures the long-term availability of hardware and software components for installed TELEPERM XS applications already installed in the plants. The further development of platform and components continues to be based on the robust, service-proven TELEPERM XS architecture, with the aim of minimizing the risks associated with equipment qualification and project licensing. A further development feature is the completion and extension of TELEPERM XS applications. This continuous innovation process, combined with maximized compatibility, makes TELEPERM XS unique, and provides the basis for a sustainable system with a service life guaranteed for the long term. Within the past 10 years, the majority of all comprehensive modernization projects worldwide were implemented or are contracted using TELEPERM XS. TELEPERM XS has been implemented in two new nuclear power plants and there are orders for four more

  20. Safety evaluation for communication network software modifications of PCS in Ulchin NPP unit 3

    International Nuclear Information System (INIS)

    Ji, S. H.; Koh, J. S.; Kim, B. R.; Oh, S. H.

    1999-01-01

    On February 2, 1999, an incident occurred at the Ulchin Nuclear Power Plant Unit 3 which resulted in the corruption of data on Perform Net of Plant Control System. This incident was caused by the ASIC (Application Specific Integrated Circuit) chip on the Rehostable Module which is a part of Network Interface Module. Regarding this incident, we required that the utility should propose new algorithms to detect the hardware failure of ASIC chip and evaluated the appropriateness of network software modifications. As a result of this evaluation process, we required that the safety related interlock signals using data communication path be hardwired to make up for the vulnerability of the system architecture. In this paper, we will discuss the system architecture of PCS and fault analysis and evaluation findings

  1. An Evaluation Method for Team Competencies to Enhance Nuclear Safety Culture

    International Nuclear Information System (INIS)

    Hang, S. M.; Seong, P. H.; Kim, A. R.

    2016-01-01

    Safety culture has received attention in safety-critical industries, including nuclear power plants (NPPs), due to various prominent accidents such as concealment of a Station Blackout (SBO) of Kori NPP unit 1 in 2012, the Sewol ferry accident in 2014, and the Chernobyl accident in 1986. Analysis reports have pointed out that one of the major contributors to the cause of the accidents is ‘the lack of safety culture’. The term, nuclear safety culture, was firstly defined after the Chernobyl accident by the IAEA in INSAG report no. 4, as follows “Safety culture is that assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, nuclear plant safety issues receive the attention warranted their significance.” Afterwards, a wide consensus grew among researchers and nuclear-related organizations, that safety culture should be evaluated and managed in a certain manner. Consequently, each nuclear-related organization defined and developed their own safety culture definitions and assessment methods. However, none of these methods provides a way for an individual or a team to enhance the safety culture of an organization. Especially for a team, which is the smallest working unit in NPPs, team members easily overlook their required practices to improve nuclear safety culture. Therefore in this study, we suggested a method to estimate nuclear safety culture of a team, by approaching with the ‘competency’ point of view. The competency is commonly focused on individuals, and defined as, “underlying characteristics of an individual that are causally related to effective or superior performance in a job.” Similar to safety culture, the definition of competency focuses on characteristics and attitudes of individuals. Thus, we defined ‘safety culture competency’ as “underlying characteristics and outward attitudes of individuals that are causally related to a healthy and strong nuclear safety

  2. Development of reliability-based safety enhancement technology

    International Nuclear Information System (INIS)

    Kim, Kil Yoo; Han, Sang Hoon; Jang, Seung Cherl

    2002-04-01

    This project aims to develop critical technologies and the necessary reliability DB for maximizing the economics in the NPP operation with keeping the safety using the information of the risk (or reliability). For the research goal, firstly the four critical technologies(Risk Informed Tech. Spec. Optimization, Risk Informed Inservice Testing, On-line Maintenance, Maintenance Rule) for RIR and A have been developed. Secondly, KIND (Korea Information System for Nuclear Reliability Data) has been developed. Using KIND, YGN 3,4 and UCN 3,4 component reliability DB have been established. A reactor trip history DB for all NPP in Korea also has been developed and analyzed. Finally, a detailed reliability analysis of RPS/ESFAS for KNSP has been performed. With the result of the analysis, the sensitivity analysis also has been performed to optimize the AOT/STI of tech. spec. A statistical analysis procedure and computer code have been developed for the set point drift analysis

  3. The bases for optimisation of scheduled repairs and tests of safety systems to improve the NPP productive efficiency

    International Nuclear Information System (INIS)

    Bilej, D.V.; Vasil'chenko, S.V.; Vlasenko, N.I.; Vasil'chenko, V.N.; Skalozubov, V.I.

    2004-01-01

    In the frames of risk-informed approaches the paper proposed the theoretical bases for methods of optimisation of scheduled repairs and tests of safety systems at nuclear power plants. The optimisation criterion is the objective risk function minimising. This function depends on the scheduled repairs/tests periodicity and the allowed time to bring the system channel to a state of non-operability. The main optimisation direct is to reduce the repair time with the purpose of enhancement of productive efficiency

  4. Development and verification of a leningrad NPP unit 1 living PSA model in the INL SAPHIRE code format for prompt operational safety level monitoring

    International Nuclear Information System (INIS)

    Bronislav, Vinnikov

    2007-01-01

    The first part of the paper presents results of the work, that was carried out in complete conformity with the Technical Assignment, which was developed by the Leningrad Nuclear Power Plant. The initial scientific and technical information, contained into the In-Depth Safety Assessment Reports, was given to the author of the work. This information included graphical Fault Trees of Safety Systems and Auxiliary Technical Systems, Event Trees for the necessary number of Initial Events, and also information about failure probabilities of basic components of the nuclear unit. On the basis of this information and fueling it to the Usa Idaho National Laboratory (INL) SAPHIRE code, we have developed an electronic version of the Data Base for failure probabilities of the components of technical systems. Then, we have developed both the electronic versions of the necessary Fault Trees, and an electronic versions of the necessary Event Trees. And at last, we have carried out the linkage of the Event Trees. This work has resulted in the Living PSA (LPSA - Living Probabilistic Safety Assessment) Model of the Leningrad NPP Unit 1. The LPSA-model is completely adapted to be consistent with the USA INL SAPHIRE Risk Monitor. The second part of the paper results in analysis of fire consequences in various places of Leningrad NPP Unit 1. The computations were carried out with the help of the LPSA-model, developed in SAPHIRE code format. On the basis of the computations the order of priority of implementation of fire prevention measures was established. (author)

  5. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    International Nuclear Information System (INIS)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B.

    1997-01-01

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.)

  6. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

    1997-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  7. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y N; Vasilenko, V A; Zasukha, V K; Migrov, Y A; Khabensky, V B [Research Inst. of Technology NITI (Russian Federation)

    1998-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  8. Dukovany nuclear power plant safety

    International Nuclear Information System (INIS)

    1999-01-01

    Presentation covers recommended safety issues for the Dukovany NPP which have been solved with satisfactory conclusions. Safety issues concerned include: radiation safety; nuclear safety; security; emergency preparedness; health protection at work; fire protection; environmental protection; chemical safety; technical safety. Quality assurance programs at all stages on NPP life time is described. Report includes description of NPP staff training provision, training simulator, emergency operating procedures, emergency preparedness, Year 2000 problem, inspections and life time management. Description of Dukovany Plant Safety Analysis Projects including integrity of the equipment, modernisation, equipment innovation and safety upgrading program show that this approach corresponds to the actual practice applied in EU countries, and fulfilment of current IAEA requirements for safety enhancement of the WWER 440/213 units in the course of MORAWA Equipment Upgrading program

  9. New developments enhancing MCNP for criticality safety

    International Nuclear Information System (INIS)

    Hendricks, J.S.; McKinney, G.W.; Forster, R.A.

    1993-01-01

    Since the early 80's MCNP has had three estimates of k eff : collision, absorption, and track length. MCNP has also had collision and absorption estimators of removal lifetime. These are calculated for every cycle and are averaged over the cycles as simple averages and covariance weighted averages. Correlation coefficients between estimators are also calculated. These criticality estimators are all in addition to the extensive summary information and tally edits used in shielding and other problems. A number of significant new developments have been made to enhance the MCNP Monte Carlo radiation transport code for criticality safety applications. These are available in the newly released MCNP4A version of the code

  10. Methods and Effects of Safety Enhancement in Korean PSR

    International Nuclear Information System (INIS)

    Kim, Young Gab; Park, Jong Woon

    2009-01-01

    Periodic Safety Review (PSR) is a comprehensive study on a nuclear power plant safety, taking into account aspects such as operational history, ageing, safety analyses and advances in code and standards since the time of construction. In Korea, PSRs have been performed for 20 units and have been effectively used to obtain an overall view of actual plant safety to determine reasonable and practical modifications that should be made in order to obtain a higher level of safety approaching that of modern plants. Among many safety enhancements achieved from Korean PSRs, new safety analyses are the important methods to confirm plant safety by increasing safety margin for specific safety issues. Methods and effects of safety enhancements applied in Korean PSRs are reviewed in this paper in light of new safety analyses to obtain additional safety margins

  11. On safety enhancements for medical robots

    International Nuclear Information System (INIS)

    Ng, W.S.; Tan, C.K.

    1996-01-01

    Both software and hardware methods to enhance safety are discussed for active medical robots applied to, among others, neurosurgery, orthopaedic surgery and prostatectomy. This paper advocates that while it is practically difficult, if not impossible, for software reliability to be 100%, there are positive measures by which a medical robot system can be made adequately or inherently safe. Such measures avoid the problems of software reliability but turn to mathematical logic directly to build a safer system. Examples in a newly developed prototype, known as surgeon assistant robot for selected urological disorders (SARUD), are given to illustrate the concept. Although software measures to promote reliability of a system is less preferred compared to hardware measures, as it can never escape from operating on a hardware platform, it is suggested that a complementary/ hybrid approach can be a good solution for achieving a safe and flexible (by being reprogrammable) system. A totally independent safety monitor is being built. It can arrest a servo runaway and detect out-of-safe-boundary conditions, using encoder pulses as input. This dedicated system can resolve some major safety concerns of a medical robot such as SARUD

  12. Software diversity: way to enhance safety?

    International Nuclear Information System (INIS)

    Dahll, G.; Bishop, P.

    1990-01-01

    The topic of the paper is the use of diversely produced programs to enhance the safety of computer-based systems applied in safety-critical areas. The paper starts with a survey of scientific investigations on the impact of software redundancy made at various institutions around the world. Main emphasis will, however, be put on the PODS/STEM projects, which have been performed at the OECD Halden Project in cooperation with the Technical Research Center of Finland, the Safety and Reliability Directorate, AEA Technology, UK, and Central Electricity Research Laboratory (now National Power Technology and Environment Centre), UK. In these projects, three program versions were made independently by three different teams, all based on the same specification. The three programs were tested back-to-back with a large amount of test data. The experience and results from this process were carefully logged and used for further analysis. Various strategies for test data selection were compared, with respect to fault finding strategies, as well as to branch and statement coverages of the tested programs. The assumption of independence of failures in diversely produced programs was investigated. A particularly interesting effect, namely failure masking due to program structure, was revealed. Static analysis techniques, software measures, and software reliability estimates were also studied. (author)

  13. Main features of buildings and structures important to safety of units V1 and V2 of Bohunice NPP

    International Nuclear Information System (INIS)

    David, M.

    1993-01-01

    The program of seismic upgrading of Bohunice NPPs has been started in the year 1989 (after finishing of new seismic input). Since that time the seismic upgrading of Main building of NPP V1 has already been realized, structural as well as technological parts. Beside that the designs of seismic upgrading of other structures of NPP V1 and V2 have been completed. It has been proved that the seismic upgrading of NPPs with reactors WWER 440 is very complicated, but still possible, even in the case with high seismic intensity. It would be not possible to fulfill this complicated task without the help of IAEA Missions. The activities of IAEA experts in the program of Bohunice NPPs upgrading are appreciated very much

  14. Selection of NPP for Kazakhstan

    International Nuclear Information System (INIS)

    Zhotabaev, Zh.R.

    2003-01-01

    Commercial NPP for Kazakhstan should to meet to several main requirements: 1). Safety operation (accident probability not more than 10 -6 1/p. year). 2). High efficiency > 40 %. 3). Possibility of use for high-temperature chemistry and hydrogen production. 4). Possibility for manufacturing of considerable part of equipment in Kazakhstan. 5). Possibility for fuel production and reprocessing in Kazakhstan. 6). Independence from existence of large water-supply sources. Comparative analysis of several NPP with different reactors (WWR-1000, Candu, BREST, VG-400; graphite molten salt reactor) shows that NPP with the graphite molten salt reactor meets to all above requirements, but hydrogen production it is possible by more complete 4-stage technology, since coolant temperature is 800 Deg. C. The principle advantage is possibility of manufacturing of main equipment and fuel in Kazakhstan that reduce the cost of NPP construction and operation

  15. Cernavoda NPP Knowledge Transfer

    International Nuclear Information System (INIS)

    Valache, C. M.

    2016-01-01

    Full text: The paper presents a description of the Knowledge Transfer (KT) process implemented at Cernavoda NPP, its designing and implementation. It is underlined that applying a KT approach should improve the value of existing processes of the organization through: • Identifying business, operational and safety risks due to knowledge gaps, • Transfer of knowledge from the ageing workforce to the peers and/or the organization, • Continually learning from successes and failures of individual or teams, • Convert tacit knowledge to explicit knowledge, • Improving operational and safety performance through creating both new knowledge and better access to existing knowledge. (author

  16. Nuclear safety analysis for transport cask TK-6 (for WWER-440) and cover for fresh assemblies (for WWER-1000) in implementation of new fuel types at Ukrainian NPP

    International Nuclear Information System (INIS)

    Bilodid, Y.; Kovbasenko, Iu; Dudka, Olena

    2006-01-01

    According to the fresh fuel management procedure, fuel assemblies - after nuclear fuel delivery to the NPP fresh fuel unit - are vertically loaded into a cover intended for the delivery of fuel assemblies into the containment of the NPP reactor compartment. The cover is placed into an universal jack in the cooling and refueling pond, and then the fresh fuel assemblies are loaded into the reactor core. Based on the nuclear safety analysis carried out by the Russian Research Center 'Kurchatov Institute' for contemporary WWER-1000 fuel, it has become necessary to limit the number of fuel assemblies loaded into a cover below its designed capacity (12 FA instead of 18 FA as originally designed). Such a decision leads to worse economic performances in fuel transportation. The paper considers potential ways to overcome this restriction. Transport container TK-6 for spent fuel assemblies was designed quite a long time ago and, as shown in this paper, the requirement on the maximally permissible neutron multiplication factor of the loaded container for individual states to be analyzed in compliance with Ukrainian regulations is not met. First of all, this concerns the container criticality analysis in optimal neutron slow-down (container filling with water-air mixture with optimal density). The paper shows potential ways for TK-6 burnup-credit loading with the maximum number of fuel assemblies and partial container loading (Authors)

  17. Bohunice V1 NPP upgrading programme

    International Nuclear Information System (INIS)

    Kerak, J.

    2001-01-01

    The paper describes whole process of Bohunice V1 NPP nuclear safety and operational reliability level increase which has been performed since units commissioning (1. unit in 1978, 2. unit in 1980), continued Small Reconstruction (1991 -1993) and finished Gradual Upgrading(1994 -2000). The main purpose is to last stage -Gradual upgrading of Bohunice V1 NPP. (author)

  18. New approach of second Romanian NPP siting

    International Nuclear Information System (INIS)

    Mauna, Traian

    2010-01-01

    The NPP sitting studies in Romania began before 1975. The first Romanian NPP CANDU 6 type reactor gone to erection in 1980 on Cernavoda site planned to have 5 units. Gained the experience from Cernavoda NPP sitting, the first mission of new multi-branch of specialists team was to choose new NPP sites adapting the NPP Cernavoda project to the new parameters of close water cooling circuit and hard less and no rock foundation strata. The studies were carrying out in different stages on the inner rivers Olt, Mures, Somes in Transylvania historical region. This paper tries to reconsider shortly the old analysis according to the last IAEA Safety Standards, taking into account the new NPP generation requirement. Paper is focused on geological aspects and other local sites characteristics. (authors)

  19. Supercompaction of radioactive waste at NPP Krsko

    International Nuclear Information System (INIS)

    Fink, K.; Sirola, P.

    1996-01-01

    The problem of radioactive waste management is both scientifically and technically complex and also deeply emotional issue. In the last twenty years the first two aspects have been mostly resolved up to the point of safe implementation. In the Republic of Slovenia, certain fundamentalist approaches in politics and the use of radioactive waste problem as a political tool, brought the final radioactive repository siting effort to a stop. Although small amounts of radioactive waste are produced in research institutes, hospitals and industry, major source of radioactive waste in Slovenia is the Nuclear Power Plant Krsko. When Krsko NPP was originally built, plans were made to construct a permanent radioactive waste disposal facility. This facility was supposed to be available to receive waste from the plant long before the on site storage facility was full. However, the permanent disposal facility is not yet available, and it became necessary to retain the wastes produced at the plant in the on-site storage facility for an extended period of time. Temporary radioactive storage capacity at the plant site has limited capacity and having no other options available NPP Krsko is undertaking major efforts to reduce waste volume generated to allow normal operation. This article describes the Radioactive Waste Compaction Campaign performed from November, 1994 through November, 1995 at Krsko NPP, to enhance the efficiency and safety of storage of radioactive waste. The campaign involved the retrieval, segmented gamma-spectrum measurement, dose rate measurement, compaction, re-packaging, and systematic storage of radioactive wastes which had been stored in the NPP radioactive waste storage building since plant commissioning. (author)

  20. Life management and operational experience feedback - tools to enhance safety and reliability of the NPP

    International Nuclear Information System (INIS)

    Mach, P.

    1997-01-01

    Preparation has started of the Temelin power plant centralized equipment database. Principles of reliability centered maintenance are studied, and use of these activities will be made in the Plant Ageing Management Programme. The aims of the Programme are as follows: selection of important components subject to ageing, data collection, determination of dominant stressors, development, selection and validation of ageing evaluation methods, setup of experience feedback, determination of responsibilities, methodologies and strategy, elaboration of programme procedures and documentation, and maintenance of programme flexibility. Pilot studies of component ageing are under way: for the reactor pressure vessel, steam generator, pressurizer, piping, ECCS and cables. The organizational structure of the Operational Experience Feedback system is described, as are the responsibility of staff and sources of information. (M.D.)

  1. IAEA Says Finland's Loviisa Nuclear Power Plant Committed to Safety, Sees Areas for Enhancement

    International Nuclear Information System (INIS)

    2018-01-01

    An International Atomic Energy Agency (IAEA) team of experts said the operator of Finland’s Loviisa Nuclear Power Plant (NPP) demonstrated a commitment to safety. The team also identified areas for further enhancement. The Operational Safety Review Team (OSART) concluded an 18-day mission on 22 March to Loviisa NPP, whose two 531-MWe pressurized-water reactors started commercial operation in 1977 and 1980, respectively. Fortum Power and Heat OY operate the plant, located about 100 km east of Helsinki, the capital. Nuclear power generates one-third of electricity in Finland, which has four operating power reactors and is constructing a fifth reactor. OSART missions aim to improve operational safety by objectively assessing safety performance using the IAEA’s safety standards and proposing improvement where appropriate. The 16-member team comprised experts from Brazil, Canada, China, France, Germany, Hungary, Romania, Russia Federation, Slovak Republic, South Africa, Spain, Ukraine, United Kingdom, United States of America as well as IAEA officials. The review covered the areas of leadership and management for safety; training and qualification; operations; maintenance; technical support; operating experience; radiation protection; chemistry; emergency preparedness and response; accident management; human, technology and organizational interactions; and long-term operation. The team identified a number of good practices that will be shared with the nuclear industry globally, including: • The plant has developed the capability to automatically calculate leak rate tests of containment. • The plant established a process to test and improve modifications and updates early. • The plant has adopted a key system to effectively control access to various rooms in the plant. The mission made several proposals to improve operational safety, including: • The plant management should improve communications of their expectations and consistently reinforce their

  2. A study on the methodology of integrated safety assessment on low and intermediate level waste (LILW) managed in temporary storage facility at NPP

    International Nuclear Information System (INIS)

    Ahn, Min Ho

    2010-02-01

    Since 1978, the KHNP has been operating 20 NPPs (16 PWRs and 4 CANDUs) and generating about 67,000 drums (200 L) of LILW (as of December 31, 2005), which have been stored in the temporary storage facility (TSF) at each NPP due to the absence of a repository for the disposal of LILW. Therefore, the period of temporary storage of LILW is so long compared to other countries. Furthermore, the details with respect to the safety analyse on the TSF have not been considered in PSAR and FSAR. Especially, the risk assessment on the TSF has scarcely been conducted as opposed to many researches on the disposal of LILW. Since 2003, however, the IAEA has been recognized on the importance of predispoal management of LILW. And then, the regulatory frame of U.S. NRC was being shifted to risk-based regulation from the deterministic approach. Therefore, most of radioactive wastes including the LILW will be managed in terms of the risk-based graded approach to future regulation system called RIR (risk informed regulation). If the radioactive wastes do not quantitatively deal with the risk-based regulation, the radiological risk on some of radioactive wastes might be overestimated or underestimated regardless of the degree of the risk. According to a consequence of these situations, the numbers of the researches on the predisposal management of LILW have been required for the preparation on new regulatory frame. In this study, the main objective of this study is to establish the methodology of integrated safety assessment on LILW managed in the TSF at NPP, and to develop the integrated safety assessment code for routine operating condition and for for accident analysis on LILW managed in the TSF. In order to establish the methodology of integrated safety assessment on LILW managed in the TSF at NPP, three main parameters were considered: risk-based accident scenarios, radionuclide inventory, and atmospheric dispersion factor (χ/Q). Arbitrary accidents related to LILW management in the

  3. Qualification of NPP operations personnel

    International Nuclear Information System (INIS)

    Wang Jiao.

    1987-01-01

    Competence of personnel is one of the important problems for safety operation of nuclear power plant. This paper gives a description of some aspects, such as the administration of NPP, posts, competence of personnel, training, assessing the competence and personnel management

  4. Dukovany NPP operation

    International Nuclear Information System (INIS)

    Vlcek, Jaroslav

    2010-01-01

    The topics discussed include: Dukovany NPP among CEZ Group power plants; International missions at the plant; Plant operation results; and Strategic goals and challenges. Historical data are presented in the graphical form, such as the unit capacity factor, unplanned capability loss factor, unplanned automatic scrams, fuel reliability, industrial safety accident rate, collective radiation exposure, WANO index, power generation data, and maximum achievable power by the end of year. Also discussed were the company culture and human resources, maintenance, power uprate, and related phenomena. (P.A.)

  5. Ongoing enhancements in the German nuclear regulatory framework with respect to fire safety

    Energy Technology Data Exchange (ETDEWEB)

    Elsche, Bjoern [e.on Kernkraft, Hannover (Germany); Roewekamp, Marina [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Koeln (Germany); Neugebauer, Wilfried [AREVA NP, Erlangen (Germany); Gersinska, Rainer [Bundesamt fuer Strahlenschutz (BfS), Salzgitter (Germany). KTA-Geschaeftsstelle

    2015-12-15

    In the recent past, the regulatory framework for nuclear power plants (NPP) in Germany has been updated and enhanced comprising on the one hand comprehensive high level regulatory documents such as the 'Safety Requirements for Nuclear Power Plants' and, on the other hand, revised state-of-the-art nuclear safety standards and rules being incorporated in a corresponding legal structure. A major enhancement concerns the nuclear fire and explosion protection standards being already available as so-called green print for final comments which are expected to be officially published end of 2015. The update became necessary after approx. ten years for better addressing some lessons learnt form the operating experience, the consideration of post- Fukushima insights, such as more systematically addressing event combinations with fires and taking into account deviations from non-nuclear standards for escape and rescue routes. Moreover, fire protections remains an important issue for nuclear power plants in Germany during the longer term post-commercial safe shutdown period before decommissioning during which the spent fuel elements remain either in the containment or in the spent fuel pool for further years requiring suitable fire protection means being in place.

  6. ASSET experience at Paks NPP

    International Nuclear Information System (INIS)

    Szabo, I.

    1997-01-01

    At Paks NPP special attention has been paid to international reviews since the very beginning of operation. Several international teams visited Paks in order to provide independent assessment of plant performance, conditions and safety. Paks NPP Management has the further intention to invite international reviews regularly (yearly) in the future as well. The experience gained during these reviews helped to establish a unified process of preparation for the reviews, performing them and handling the results. The Safety Department is in charge of organization of the whole process. All these reviews have their specific features and they are focused on different areas. The ASSET reviews provides the assessment of plant performance and safety through the analysis of safety significant events, which have occurred at the nuclear power plant. This approach makes this review specific and different from the other ones

  7. ASSET experience at Paks NPP

    Energy Technology Data Exchange (ETDEWEB)

    Szabo, I [Operational Safety Dept., Paks NPP, Paks (Hungary)

    1997-10-01

    At Paks NPP special attention has been paid to international reviews since the very beginning of operation. Several international teams visited Paks in order to provide independent assessment of plant performance, conditions and safety. Paks NPP Management has the further intention to invite international reviews regularly (yearly) in the future as well. The experience gained during these reviews helped to establish a unified process of preparation for the reviews, performing them and handling the results. The Safety Department is in charge of organization of the whole process. All these reviews have their specific features and they are focused on different areas. The ASSET reviews provides the assessment of plant performance and safety through the analysis of safety significant events, which have occurred at the nuclear power plant. This approach makes this review specific and different from the other ones.

  8. Enhancement of safety for reprocessing facilities

    International Nuclear Information System (INIS)

    2012-06-01

    The adequacy of the safety measures for utility loss accidents in nuclear fuel reprocessing facilities which have been formulated by the nuclear enterprises is investigated in JNES which organizes an advanced committee to specifically study this problem. The results are reviewed in the present report including the case of such severe accidents as in Fukushima Daiichi Nuclear Power Plant. The report also represents a tentative proposal for examination standards of such unimaginable severe accidents as 'station blackout,' urgent safety measures necessary for reoperation of nuclear power plants and requested by nuclear and industrial safety agency, and pointing out and clarification of the potential weakness from the safety point of view, and collective and composite evaluation of safety of the relevant facilities. Furthermore, the definition of accident management is given as of controlled condition and the authorized way of thinking for the cases of plural events happening at the same time and the cases when risks exist radioactivity emits with explosion. (S. Ohno)

  9. Application of ant colony optimization in NPP classification fault location

    International Nuclear Information System (INIS)

    Xie Chunli; Liu Yongkuo; Xia Hong

    2009-01-01

    Nuclear Power Plant is a highly complex structural system with high safety requirements. Fault location appears to be particularly important to enhance its safety. Ant Colony Optimization is a new type of optimization algorithm, which is used in the fault location and classification of nuclear power plants in this paper. Taking the main coolant system of the first loop as the study object, using VB6.0 programming technology, the NPP fault location system is designed, and is tested against the related data in the literature. Test results show that the ant colony optimization can be used in the accurate classification fault location in the nuclear power plants. (authors)

  10. Spain in South Ukraine NPP

    International Nuclear Information System (INIS)

    Ibanez, M.

    1994-01-01

    A Technical Assistance Protocol was signed between the Governments of the GIS and the Commission of the European Union (CEU) on August 2, 1991 and this was the starting point of the TACIS program. In this article, the activities described are those related to the TACIS-92/93/94 on site technical assistance to South Ukraine NPP (SUK NPP). Within the scope of the TACIS 92 Program the CEU and the Ukrainian Authorities agreed a list of projects to be implemented at South Ukraine NPP with the aim to improve the operational safety of the plant. This part of the program is called TACIS 92 on-site activities. The total budget allocated to these projects is a MECU. The European Union ''utility'' selected to lead this program at South Ukraine NPP was UNESA and the first contract to cover our activities was signed in July 1993 between the CEU (Mr. Pablo Benavides) and UNESA (Mr. Pedro Rivero). The projects will be implemented at SUK NPP but according to the contract UNESA is ''The Consultant'' and GOSKOMATON (The Ukrainian Sate Committee on Nuclear Power Utilization) is the ''Recipient Institution''. (Author)

  11. Safety analysis on Non-LOCA events for the revision of Wolsong NPP unit 2,3,4 sar

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Ryu, Eui Seung; Kho, Dong Wook; Kim, Sung Min

    2015-01-01

    Korean Wolsong Nuclear Power Plant Units 2,3,4 (CANDU-6 Type) has prepared the revision of safety analysis report (Final Safety Analysis Report (FSAR) chapter 15) from the original performed in the year of 1990s, using the updated and state-of-the-art methodology and tools including IST safety analysis codes and more detail modelling. Compared with the original FSAR15, the revised FSAR15 has significant improvement in both the scope and the depth of safety analysis, which has demonstrated the safety analysis results have complied with the safety requirements(acceptance criteria). This paper will present the analysis scope for Non-LOCA events re-analyzed or added for the FSAR15 revision, methodologies applied such as codes and modelling and some important analysis results will be demonstrated with comparison to acceptance criteria. Application of more detail and near-realistic assumptions and method including Dev-PDO options and uncertainty related to the CHF correlations has altogether brought about more safety margin compared with the original FSAR15 with respect to SDS trip effectiveness etc. (author)

  12. Safety enhancement through calibration and maintenance

    International Nuclear Information System (INIS)

    Mohammad Ali Khalid; Taiman Kadni

    2000-01-01

    Although radiation causes an alarming effect to living organism, it has been accepted for diagnostic and therapy in medicine as well as quality control and test method in industry. Due to the benefit gained, public at large is willing to accept the risk on condition that proper protection and safety standard is implemented and practised. Instrument plays a very important role not only in generation radiation but also measuring it. Therefore to ensure safety it is fair to say that all instrument must be calibrated to a certain standard and maintained to operate to its design specifications from time to time. (author)

  13. Decommissioning of NPP A-1

    International Nuclear Information System (INIS)

    Anon

    2009-01-01

    In this presentation the Operation history of A1 NPP, Project 'Decommissioning of A1 NPP' - I stage, Project 'Decommissioning of A1 NPP ' - II stage and Next stages of Project 'Decommissioning of A1 NPP ' are discussed.

  14. An effective technique for the software requirements analysis of NPP safety-critical systems, based on software inspection, requirements traceability, and formal specification

    International Nuclear Information System (INIS)

    Koo, Seo Ryong; Seong, Poong Hyun; Yoo, Junbeom; Cha, Sung Deok; Yoo, Yeong Jae

    2005-01-01

    A thorough requirements analysis is indispensable for developing and implementing safety-critical software systems such as nuclear power plant (NPP) software systems because a single error in the requirements can generate serious software faults. However, it is very difficult to completely analyze system requirements. In this paper, an effective technique for the software requirements analysis is suggested. For requirements verification and validation (V and V) tasks, our technique uses software inspection, requirement traceability, and formal specification with structural decomposition. Software inspection and requirements traceability analysis are widely considered the most effective software V and V methods. Although formal methods are also considered an effective V and V activity, they are difficult to use properly in the nuclear fields as well as in other fields because of their mathematical nature. In this work, we propose an integrated environment (IE) approach for requirements, which is an integrated approach that enables easy inspection by combining requirement traceability and effective use of a formal method. The paper also introduces computer-aided tools for supporting IE approach for requirements. Called the nuclear software inspection support and requirements traceability (NuSISRT), the tool incorporates software inspection, requirement traceability, and formal specification capabilities. We designed the NuSISRT to partially automate software inspection and analysis of requirement traceability. In addition, for the formal specification and analysis, we used the formal requirements specification and analysis tool for nuclear engineering (NuSRS)

  15. Lessons learnt from the resin release into the primary circuit of the Fessenheim NPP unit 1 in January 2004. Impact on the nuclear safety

    International Nuclear Information System (INIS)

    Georgescu, M.

    2004-01-01

    On January the 24 th , at the Fessenheim NPP unit 1, a human error was committed during a boron demineralizer line-up, caused by lack of preparation. Consequently, a quantity of resin estimated at about 300 liters was released from this demineralizer, through its safety valve, into the head-tank of the chemical and volume control (CVC) system and after that, into the primary circuit. The incident had a real impact on the unit: the CVC filters were clogged, the seal injection flow of the primary circuit main pumps was lost, the primary circuit main pump 2 tripped four days after the incident, as the rate of the recirculated seal leak flow (downstream the seal 1) increased up to the automatic trip set point, the shaft of the running primary circuit feed pump was found seized into the rear hydrostatic bearing following the pump stop (after ten days of successful operation), the thimble plugs were jammed into their guide tubes, the small diameter pipes were plugged. The unit shutdown for over five months was necessary to clean the primary circuit components, repair or replace the affected equipment items and carry out inspections and tests. The reinforced unit in-service monitoring program, set up during the unit start-up, confirms that, up to now, the unit operation has not been adversely affected by the residual amounts of resin which subsist in certain areas of the primary circuit. Nevertheless, it remains to verify that, in the long term, these deposits will have no negative chemical effect in the potential confined areas, such as the thermal barriers of the primary circuit main pumps. Finally, the occurrence of this incident underlines, once more, the importance of normal operating activity preparing and checking. It also reveals the implementation of an ''unforgiving'' design change allowing the installation of a boron demineralizer safety valve having its outlet connected to the primary circuit. (orig.)

  16. Enhancing Safety at Airline Operations Control Centre

    Directory of Open Access Journals (Sweden)

    Lukáš Řasa

    2015-04-01

    Full Text Available In recent years a new term of Safety Management System (SMS has been introduced into aviation legislation. This system is being adopted by airline operators. One of the groundbased actors of everyday operations is Operations Control Centre (OCC. The goal of this article has been to identify and assess risks and dangers which occur at OCC and create a template for OCC implementation into SMS.

  17. Advanced repair methods for enhanced reactor safety

    International Nuclear Information System (INIS)

    Kornfeldt, H.

    1993-01-01

    A few innovative concepts are described of the ABB Atom Service Division for repair and mitigation techniques for primary systems in nuclear power plants. The concepts are based on Shape Memory Alloy (SMA) technology. A basic feature of all methods is that welding and component replacement is being avoided and the radiation dose imposed on maintenance personnel reduced. The SMA-based repair methods give plant operators new ways to meet increased safety standards and rising maintenance costs. (Z.S.) 4 figs

  18. Mochovce NPP simulator

    International Nuclear Information System (INIS)

    Ziakova, M.

    1998-01-01

    Mochovce NPP simulator basic features and detailed description of its characteristics are presented with its performance, certification and application for training of NPP operators as well as the training scenario

  19. Physician attitudes towards pharmacological cognitive enhancement: safety concerns are paramount.

    Directory of Open Access Journals (Sweden)

    Opeyemi C Banjo

    2010-12-01

    Full Text Available The ethical dimensions of pharmacological cognitive enhancement have been widely discussed in academic circles and the popular media, but missing from the conversation have been the perspectives of physicians - key decision makers in the adoption of new technologies into medical practice. We queried primary care physicians in major urban centers in Canada and the United States with the aim of understanding their attitudes towards cognitive enhancement. Our primary hypothesis was that physicians would be more comfortable prescribing cognitive enhancers to older patients than to young adults. Physicians were presented with a hypothetical pharmaceutical cognitive enhancer that had been approved by the regulatory authorities for use in healthy adults, and was characterized as being safe, effective, and without significant adverse side effects. Respondents overwhelmingly reported increasing comfort with prescribing cognitive enhancers as the patient age increased from 25 to 65. When asked about their comfort with prescribing extant drugs that might be considered enhancements (sildenafil, modafinil, and methylphenidate or our hypothetical cognitive enhancer to a normal, healthy 40 year old, physicians were more comfortable prescribing sildenafil than any of the other three agents. When queried as to the reasons they answered as they did, the most prominent concerns physicians expressed were issues of safety that were not offset by the benefit afforded the individual, even in the face of explicit safety claims. Moreover, many physicians indicated that they viewed safety claims with considerable skepticism. It has become routine for safety to be raised and summarily dismissed as an issue in the debate over pharmacological cognitive enhancement; the observation that physicians were so skeptical in the face of explicit safety claims suggests that such a conclusion may be premature. Thus, physician attitudes suggest that greater weight be placed upon the

  20. Potential improvement of CANDU NPP safety margins by shortening the response time of shutdown systems using FPGA based implementation

    Energy Technology Data Exchange (ETDEWEB)

    Jingke She, E-mail: jshe2@uwo.ca [Department of Electrical and Computer Engineering, University of Western Ontario, London, Ontario N6A 5B9 (Canada); Jin Jiang, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, University of Western Ontario, London, Ontario N6A 5B9 (Canada)

    2012-03-15

    Highlights: Black-Right-Pointing-Pointer Quantitative analysis of the safety margin improvement through thermalhydraulic simulation and analysis. Black-Right-Pointing-Pointer Hardware-in-the-loop simulation of realizing the improvement by an FPGA-based SDS1. Black-Right-Pointing-Pointer Verification of potential operating power upgrade without endangering the plant safety. - Abstract: The relationship between the peak values of critical reactor variables, such as neutronic power, inside a CANDU reactor and the speed of the response of its shutdown system has been analyzed in the event of a large loss of coolant accident (LOCA). The advantage of shortening the response time of the shutdown action has been demonstrated in term of the improved safety margin. A field programmable gate array (FPGA) platform has been chosen to implement such a shutdown system. Hardware-in-the-loop (HIL) simulations have been performed to demonstrate the feasibility of this concept. Furthermore, connections between the speed of response of the shutdown system and the nominal operating power level of the reactor have been drawn to support for potential power upgrade for existing power plants.

  1. Electrical Grid Conditioning For First NPP Integration, a Systems Engineering Approach Incorporating Quality Function Deployment

    Energy Technology Data Exchange (ETDEWEB)

    Pwani, Henry; James, J. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2013-10-15

    Nuclear power plant has a high potential to cause serious harm to environment as evidenced by effects of Fukushima and Chernobyl accidents. A reliable electrical power is required for a NPP to facilitate cooling after a shutdown. Failure of electrical power supply during shutdown increases core damage probability. Research shows that a total of 39% of LOOP related events in US are electrical grid centered. In Korea, 38% and 29% of all events that led to NPP shutdown at Hanul units 3-6 and at Hanbit units 3-6 respectively were electrical related. Electric grids for both operating and new NPPs must therefore be examined and upgraded for reliability improvement in order to enhance NPP safety.

  2. VERONA V6.22 – An enhanced reactor analysis tool applied for continuous core parameter monitoring at Paks NPP

    Energy Technology Data Exchange (ETDEWEB)

    Végh, J., E-mail: janos.vegh@ec.europa.eu [Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Pós, I., E-mail: pos@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Horváth, Cs., E-mail: csaba.horvath@energia.mta.hu [Centre for Energy Research, Hungarian Academy of Sciences, H-1525 Budapest 114, P.O. Box 49 (Hungary); Kálya, Z., E-mail: kalyaz@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Parkó, T., E-mail: parkot@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary); Ignits, M., E-mail: ignits@npp.hu [Paks Nuclear Power Plant Ltd., H-7031 Paks, P.O. Box 71 (Hungary)

    2015-10-15

    Between 2003 and 2007 the Hungarian Paks NPP performed a large modernization project to upgrade its VERONA core monitoring system. The modernization work resulted in a state-of-the-art system that was able to support the reactor thermal power increase to 108% by more accurate and more frequent core analysis. Details of the new system are given in Végh et al. (2008), the most important improvements were as follows: complete replacement of the hardware and the local area network; application of a new operating system and porting a large fraction of the original application software to the new environment; implementation of a new human-system interface; and last but not least, introduction of new reactor physics calculations. Basic novelty of the modernized core analysis was the introduction of an on-line core-follow module based on the standard Paks NPP core design code HELIOS/C-PORCA. New calculations also provided much finer spatial resolution, both in terms of axial node numbers and within the fuel assemblies. The new system was able to calculate the fuel applied during the first phase of power increase accurately, but it was not tailored to determine the effects of burnable absorbers as gadolinium. However, in the second phase of the power increase process the application of fuel assemblies containing three fuel rods with gadolinium content was intended (in order to optimize fuel economy), therefore off-line and on-line VERONA reactor physics models had to be further modified, to be able to handle the new fuel according to the accuracy requirements. In the present paper first a brief overview of the system version (V6.0) commissioned after the first modernization step is outlined; then details of the modified off-line and on-line reactor physics calculations are described. Validation results for new modules are treated extensively, in order to illustrate the extent and complexity of the V&V procedure associated with the development and licensing of the new

  3. Assessment of the event at Rovno NPP owing to the unexpected opening and obstruction of the safety valve lock mechanism of the pressure compensator

    International Nuclear Information System (INIS)

    Alonso, C.

    1993-01-01

    The main objective of this analysis is to be able to identify the sequence of the accident and evaluate the real frequency observed in similar nuclear power plants, according to the experience registered in the data bases of the NPP accident information system (ISI-AES) as well as the quality assurance information system (ISKO). This work describes how the analysis of events in a VVER-440 reactor NPP was performed

  4. Defining safety culture and the nexus between safety goals and safety culture. 4. Enhancing Safety Culture Through the Establishment of Safety Goals

    International Nuclear Information System (INIS)

    Tateiwa, Kenji; Miyata, Koichi; Yahagi, Kimitoshi

    2001-01-01

    Safety culture is the perception of each individual and organization of a nuclear power plant that safety is the first priority, and at Tokyo Electric Power Company (TEPCO), we have been practicing it in everyday activities. On the other hand, with the demand for competitiveness of nuclear power becoming even more intense these days, we need to pursue efficient management while maintaining the safety level at the same time. Below, we discuss how to achieve compatibility between safety culture and efficient management as well as enhance safety culture. Discussion at Tepco: safety culture-nurturing activities such as the following are being implemented: 1. informing the employees of the 'Declaration of Safety Promotion' by handing out brochures and posting it on the intranet home page; 2. publishing safety culture reports covering stories on safety culture of other industry sectors, recent movements on safety culture, etc.; 3. conducting periodic questionnaires to employees to grasp how deeply safety culture is being established; 4. carrying out educational programs to learn from past cases inside and outside the nuclear industry; 5. committing to common ownership of information with the public. The current status of safety culture in Japan sometimes seems to be biased to the quest of ultimate safety; rephrasing it, there have been few discussions regarding the sufficiency of the quantitative safety level in conjunction with the safety culture. Safety culture is one of the most crucial foundations guaranteeing the plant's safety, and for example, the plant safety level evaluated by probabilistic safety assessment (PSA) could be said to be valid only on the ground that a sound and sufficient safety culture exists. Although there is no doubt that the safety culture is a fundamental and important attitude of an individual and organization that keeps safety the first priority, the safety culture in itself should not be considered an obstruction to efforts to implement

  5. Experimental studies of thermo-hydraulic processes during passive safety systems operation in new WWER NPP projects

    International Nuclear Information System (INIS)

    Morozov, A.V.; Remizov, O.V.; Kalyakin, D.S.

    2014-01-01

    The results of experimental study of thermal-hydraulic processes during operation of the passive safety systems of WWER reactors of new generation are given. The interaction processes of counter flows of saturated steam and cold water in vertical steam-line of the auxiliary passive core reflood system from secondary hydraulic accumulator are studied. The peculiarities of undeveloped boiling on single horizontal tube heating by steam and steam-gas mixture, which is character for WWER steam generator condensing mode, are investigated [ru

  6. The LBB methodology application results performed on the safety related piping of NPP V-1 in Jaslovske Bohunice

    Energy Technology Data Exchange (ETDEWEB)

    Kupca, L.; Beno, P. [Nuclear Power Plants Research Institute, Trnava (Slovakia)

    1997-04-01

    A broad overview of the leak before break (LBB) application to the Slovakian V-1 nuclear power plant is presented in the paper. LBB was applied to the primary cooling circuit and surge lines of both WWER 440 type units, and also used to assess the integrity of safety related piping in the feed water and main steam systems. Experiments and calculations performed included analyses of stresses, material mechanical properties, corrosion, fatigue damage, stability of heavy component supports, water hammer, and leak rates. A list of analysis results and recommendations are included in the paper.

  7. Brief account of the design philosophy for third Qinshan NPP shutdown safety system based on practical application

    International Nuclear Information System (INIS)

    Xiong Weihua

    2005-01-01

    Qinshan CANDU power plant is uses the Canadian proven CANDU6 nuclear power technology. It has two characteristic: 1. heavy water-as moderator and coolant; 2. natural uranium as the fuel and change fuel during normal operating. CANDU6 include four special safety system: the No.1 shutdown system (SDS No.1), the No.2 shutdown system (SDS No.2), the containment system, the emergency core cooling system (ECCS). QinShan CANDU power plant is the first commercial PHWR nuclear power plant in China. And some aspect is not similar to everybody. The intention of the article is to introduce the basic design and functions. (authors)

  8. Enhancing Safety of Artificially Ventilated Patients Using Ambient Process Analysis.

    Science.gov (United States)

    Lins, Christian; Gerka, Alexander; Lüpkes, Christian; Röhrig, Rainer; Hein, Andreas

    2018-01-01

    In this paper, we present an approach for enhancing the safety of artificially ventilated patients using ambient process analysis. We propose to use an analysis system consisting of low-cost ambient sensors such as power sensor, RGB-D sensor, passage detector, and matrix infrared temperature sensor to reduce risks for artificially ventilated patients in both home and clinical environments. We describe the system concept and our implementation and show how the system can contribute to patient safety.

  9. NPP Mochovce - a project of extraordinay significance

    International Nuclear Information System (INIS)

    Chwolik, I.; Debru, M.

    2000-01-01

    In this paper and in this presentation the reactor safety upgrading of two blocks of the NPP V-1 Bohunice, some results of participation on safety upgrading by the German-French consortium EUCOM (Framatome and Siemens-KWU) are presented. (author)

  10. On psichological problem of NPP operation and control

    International Nuclear Information System (INIS)

    Mashin, V.A.

    1994-01-01

    The role of psichological factor as a reserve for increasing NPP safety connected with human factor is discussed. It is emphasized that the process of NPP personnel professional training should not be restricted by formation of a certain set of knowledge, skills and experience. It is necessary to initiate ability for constant self-developing. Control for assurance of effective interaction of the whole NPP personnel is an important problem

  11. PNRA Process for Utilizing Experience Feedback for Enhancing Nuclear Safety

    International Nuclear Information System (INIS)

    Shah, Z.H.

    2016-01-01

    One of the elements essential for any organization to become a learning organization is to learn from its own and others experience. The importance of utilizing experience feedback for enhancing operational safety is highlighted in nuclear industry again and again and this has resulted in establishment of several national and international forums. In addition, IAEA action plan on nuclear safety issued after Fukushima accident further highlighted the importance of experience sharing among nuclear community to enhance global nuclear safety regime. PNRA utilizes operating experience feedback gathered through different sources in order to improve its regulatory processes. During the review of licensing submissions, special emphasis is given to utilize the lessons learnt from experience feedback relating to nuclear industry within and outside the country. This emphasis has gradually resulted in various safety improvements in the facilities and processes. Accordingly, PNRA has developed a systematic process of evaluation of international operating experience feedback with the aim to create safety conscious approach. This process includes collecting information from different international forums such as IAEA, regulatory bodies of other countries and useful feedback of past accidents followed by its screening, evaluation and suggesting recommendations both for PNRA and its licensees. As a result of this process, several improvements concerning regulatory inspection plans of PNRA as well as in regulatory decision making and operational practices of licensees have been highlighted. This paper will present PNRA approach for utilizing experience feedback in its regulatory processes for enhancing / improving nuclear safety. (author)

  12. Ignalina Safety Analysis Group

    International Nuclear Information System (INIS)

    Ushpuras, E.

    1995-01-01

    The article describes the fields of activities of Ignalina NPP Safety Analysis Group (ISAG) in the Lithuanian Energy Institute and overview the main achievements gained since the group establishment in 1992. The group is working under the following guidelines: in-depth analysis of the fundamental physical processes of RBMK-1500 reactors; collection, systematization and verification of the design and operational data; simulation and analysis of potential accident consequences; analysis of thermohydraulic and neutronic characteristics of the plant; provision of technical and scientific consultations to VATESI, Governmental authorities, and also international institutions, participating in various projects aiming at Ignalina NPP safety enhancement. The ISAG is performing broad scientific co-operation programs with both Eastern and Western scientific groups, supplying engineering assistance for Ignalina NPP. ISAG is also participating in the joint Lithuanian - Swedish - Russian project - Barselina, the first Probabilistic Safety Assessment (PSA) study of Ignalina NPP. The work is underway together with Maryland University (USA) for assessment of the accident confinement system for a range of breaks in the primary circuit. At present the ISAG personnel is also involved in the project under the grant from the Nuclear Safety Account, administered by the European Bank for reconstruction and development for the preparation and review of an in-depth safety assessment of the Ignalina plant

  13. Experience gained in enhancing operational safety at ComEd`s nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Elias, D [Commonwealth Edison Co. (United States)

    1997-09-01

    The following aspects of experience gained in enhancing operational safety at Comed`s nuclear power plants are discussed: nuclear safety policy; centralization/decentralization; typical nuclear operating organization; safety review boards; human performance enhancement; elements of effective nuclear oversight.

  14. Experience gained in enhancing operational safety at ComEd's nuclear power plants

    International Nuclear Information System (INIS)

    Elias, D.

    1997-01-01

    The following aspects of experience gained in enhancing operational safety at Comed's nuclear power plants are discussed: nuclear safety policy; centralization/decentralization; typical nuclear operating organization; safety review boards; human performance enhancement; elements of effective nuclear oversight

  15. Safety culture in nuclear installations. Guidance for the use in enhancement of safety culture

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-12-01

    IAEA Assessment of the Safety Culture in Organizations Team (ASCOT) activities, significant work has been done with respect to indicators to determine the effectiveness of safety culture. IAEA-TECDOC-860, ASCOT Guidelines, issued in 1996, proposed key indicators for the different areas that need to be considered when assessing safety culture. A number of other INSAG reports and Safety Series reports have been published by the IAEA to provide information that will be helpful to organizations that are interested in developing or enhancing their safety culture. Recently the IAEA perspective on safety culture has broadened even further with attention focused on obtaining a deeper understanding of the actual concept of culture and particularly organizational culture. The approach supplements the IAEA previous efforts to promote a better understanding of safety culture, but does not replace them. It can be argued that the IAEA perspective of safety culture has evolved in stages. The first stage was associated with the definition of safety culture; the second stage with how to assess it; the third stage on how to enhance it; and the fourth and current stage on obtaining a deeper understanding of the actual concept of culture. The hope is that the more sophisticated understanding obtained in the fourth stage will lead to improved ways of assessing safety culture, and also increase our ability to detect changes in safety culture at an early point in time. This ability will enable corrective actions to be taken to avoid the consequences of a deteriorating safety culture.

  16. Safety culture in nuclear installations. Guidance for the use in enhancement of safety culture

    International Nuclear Information System (INIS)

    2002-12-01

    IAEA Assessment of the Safety Culture in Organizations Team (ASCOT) activities, significant work has been done with respect to indicators to determine the effectiveness of safety culture. IAEA-TECDOC-860, ASCOT Guidelines, issued in 1996, proposed key indicators for the different areas that need to be considered when assessing safety culture. A number of other INSAG reports and Safety Series reports have been published by the IAEA to provide information that will be helpful to organizations that are interested in developing or enhancing their safety culture. Recently the IAEA perspective on safety culture has broadened even further with attention focused on obtaining a deeper understanding of the actual concept of culture and particularly organizational culture. The approach supplements the IAEA previous efforts to promote a better understanding of safety culture, but does not replace them. It can be argued that the IAEA perspective of safety culture has evolved in stages. The first stage was associated with the definition of safety culture; the second stage with how to assess it; the third stage on how to enhance it; and the fourth and current stage on obtaining a deeper understanding of the actual concept of culture. The hope is that the more sophisticated understanding obtained in the fourth stage will lead to improved ways of assessing safety culture, and also increase our ability to detect changes in safety culture at an early point in time. This ability will enable corrective actions to be taken to avoid the consequences of a deteriorating safety culture

  17. New enhancements to SCALE for criticality safety analysis

    International Nuclear Information System (INIS)

    Hollenbach, D.F.; Bowman, S.M.; Petrie, L.M.; Parks, C.V.

    1995-01-01

    As the speed, available memory, and reliability of computer hardware increases and the cost decreases, the complexity and usability of computer software will increase, taking advantage of the new hardware capabilities. Computer programs today must be more flexible and user friendly than those of the past. Within available resources, the SCALE staff at Oak Ridge National Laboratory (ORNL) is committed to upgrading its computer codes to keep pace with the current level of technology. This paper examines recent additions and enhancements to the criticality safety analysis sections of the SCALE code package. These recent additions and enhancements made to SCALE can be divided into nine categories: (1) new analytical computer codes, (2) new cross-section libraries, (3) new criticality search sequences, (4) enhanced graphical capabilities, (5) additional KENO enhancements, (6) enhanced resonance processing capabilities, (7) enhanced material information processing capabilities, (8) portability of the SCALE code package, and (9) other minor enhancements, modifications, and corrections to SCALE. Each of these additions and enhancements to the criticality safety analysis capabilities of the SCALE code system are discussed below

  18. Licensing of the Ignalina NPP

    International Nuclear Information System (INIS)

    Kutas, S.

    1999-01-01

    Since 1991 State Nuclear Power Safety Inspectorate (VATESI) has regulated Ignalina NPP operation by issuing annual operating permits. Those have been issued following submission of specified documents by the Ignalina NPP that have been reviewed by VATESI. However, according to to the procedures that are now established in the Law on Nuclear Energy and subordinate regulations the use of nuclear energy in the Republic of Lithuania is subject to strict licensing. Therefore a decision about the licence for continued operation of unit 1 should be taken. Licence would be granted by VATESI in cooperation with the Ministry of Health, Ministry of Environment and the institutions of local authorities. Ignalina NPP presented to the VATESI safety analysis report (SAR) with other documents. SAR was made mainly by foreign experts and financed by European Bank for Reconstruction and Development (EBRD). VATESI in this process is supported by western regulators. A special project LAP - Licensing Assistance Project was launched to help VATESI perform licensing according western practices

  19. Knowledge management and networking for enhancing nuclear safety

    International Nuclear Information System (INIS)

    Taniguchi, T.; Lederman, L.

    2004-01-01

    Striving for innovative solutions to enhance efficiency of programme delivery and a wider outreach of its nuclear safety activities, the International Atomic Energy Agency (IAEA) has developed an Integrated Safety Approach as a platform for linking its safety related statutory functions and its many associated activities. The approach recognizes the vital importance of effective management of the knowledge base and builds on the integration between the IAEA's safety standards and all aspects of the provision for their application, including peer reviews and technical meetings to share lessons learned. The IAEA is using knowledge management techniques to develop process flows, map safety knowledge and to promote knowledge sharing. The first practical application was the establishment of a knowledge base related to safety aspects of ageing and long-term operation of nuclear power plants. The IAEA is also promoting and facilitating the establishment of regional nuclear and radiation safety networks to preserve existing knowledge and expertise as well as to strengthen sharing and creation of new knowledge in these fields. Prominent examples are the Asian Nuclear Safety Network established in the frame of the IAEA's Programme on the Safety of Nuclear Installations in South East Asia, Pacific and Far East Countries, and the Ibero-American Radiation Safety Network in the frame of the Ibero-American Forum of Nuclear Regulators. Results to date are most encouraging and suggest that this pioneer work should be extended to other regions and eventually to a global nuclear safety network. Responsive to the need of Member States, the IAEA Secretariat has prepared and made available a large number of up-to-date training packages in nuclear, radiation, transport and waste safety, using IAEA safety standards as a basis. It is also providing instruction to trainers in Member States on the use of these modules. This ensures that the material is properly used and that the IAEA

  20. Ukrainian Nuclear Society International Conference 'Modernization of the NPP with VVER reactor' (abstracts)

    International Nuclear Information System (INIS)

    Bar'yakhtar, V.G.

    1999-01-01

    Abstracts of the papers presented at International conference of the Ukrainian Nuclear Society 'Modernization of the NPP with VVER reactor'. The following problems are considered: improving the NPP's safety and reliability; reactor modernization, the lifetime prolongation; increasing of the reactor operating characteristics; methods of capacity factor increasing: refueling control, maintenance control; technical and economical aspects of NPP modernization; modernization of the automated control system of the fuel process at the NPP's; technical features and methods for the continued radiation and technology control at the NPP's; training, increasing the staff qualification and NPP modernization

  1. Decommissioning of Brennilis NPP

    International Nuclear Information System (INIS)

    Baize, Jean-Marc

    1998-01-01

    This EDF press communique give information related to the decommissioning of the Brennilis NPP. The following five items are developed in this report: 1. the level-2 decommissioning operations at the Brennilis NPP; 2. the Brennilis NPP, a pilot operation from the commissioning up to the decommissioning; 3. history of the Brennilis NPP decommissioning; 4. the types of radioactive wastes generated by the Brennilis NPP decommissioning; 5. the Brennilis NPP - a yard management as a function of the wastes. The document contains also seven appendices addressing the following subjects: 1. the share of decommissioning assigned to EDF and the decommissioning steps; 2. the EDF installations in course of decommissioning; 3. the CEA decommissioned installations or in course of decommissioning; 4. regulations; 5. costs; 6. waste management - principles; 7. data on the decommissioning yard

  2. Using the digital reactor control systems at NPP

    International Nuclear Information System (INIS)

    Schirl, G.; Hertel, J.

    2006-01-01

    A conception of application of the digital reactor control systems (RCS) at NPP is presented. The digital RCS architecture and safety ensuring are considered. The strategy and algorithm of the operating NPP equipping with the new digital RCS are given too [ru

  3. Low-water considerations for NPP sites

    International Nuclear Information System (INIS)

    Jensen, J.; Frank, T.; Wahl, T.

    2009-01-01

    In the frame of reactor safety considerations flood protection and the evaluation of low-water events are safety relevant issues. Therefore low-water statistics were performed for the coastal region Cuxhaven and the Elbe river estuary. The consideration of the longitudinal profile of water levels during low-tide in connection with surface water effects, morphodynamic changes of the river and anthropogenic modifications is of importance for conclusions concerning the NPP sites in some distance of the reference tide gauge. The authors performed a statistical low-tide analysis for the NPP sites Brunsbuettel and Brokdorf.

  4. Near Regional and Site Investigations of the Temelin NPP Site

    International Nuclear Information System (INIS)

    Prachar, Ivan; Vacek, Jiri; Heralecky, Pavel

    2011-01-01

    The Temelin NPP is worldwide through heated discussion with nuclear energetic opposition. In addition this discussion goes beyond a border of the Czech Republic. On the other side, results of several international supervisions shown that Temelin NPP is fully comparable with the safest nuclear power plants in the world regarding its technical design and safety functions. This presentation deals with the near regional and site investigations of the Temelin NPP Site. It must be noted that although the Temelin site is situated in the area with low seismicity, item of seismicity is a basic argument against Temelin NPP and therefore a detail seismic hazard assessment was performed

  5. NPP life management (abstracts)

    International Nuclear Information System (INIS)

    Litvinskij, L.L.; Barbashev, S.V.

    2002-01-01

    Abstracts of the papers presented at the International conference of the Ukrainian Nuclear Society 'NPP Life Management'. The following problems are considered: modernization of the NPP; NPP life management; waste and spent nuclear fuel management; decommissioning issues; control systems (including radiation and ecological control systems); information and control systems; legal and regulatory framework. State nuclear regulatory control; PR in nuclear power; training of personnel; economics of nuclear power engineering

  6. Use of a web site to enhance criticality safety training

    International Nuclear Information System (INIS)

    Huang, Song T.; Morman, James A.

    2003-01-01

    Establishment of the NCSP (Nuclear Criticality Safety Program) website represents one attempt by the NCS (Nuclear Criticality Safety) community to meet the need to enhance communication and disseminate NCS information to a wider audience. With the aging work force in this important technical field, there is a common recognition of the need to capture the corporate knowledge of these people and provide an easily accessible, web-based training opportunity to those people just entering the field of criticality safety. A multimedia-based site can provide a wide range of possibilities for criticality safety training. Training modules could range from simple text-based material, similar to the NCSET (Nuclear Criticality Safety Engineer Training) modules, to interactive web-based training classes, to video lecture series. For example, the Los Alamos National Laboratory video series of interviews with pioneers of criticality safety could easily be incorporated into training modules. Obviously, the development of such a program depends largely upon the need and participation of experts who share the same vision and enthusiasm of training the next generation of criticality safety engineers. The NCSP website is just one example of the potential benefits that web-based training can offer. You are encouraged to browse the NCSP website at http://ncsp.llnl.gov. We solicit your ideas in the training of future NCS engineers and welcome your participation with us in developing future multimedia training modules. (author)

  7. Enhancing Road Safety Behaviour Using a Psychological and Spiritual Approaches

    Directory of Open Access Journals (Sweden)

    Ghous Mohd Tarmizi

    2017-01-01

    Full Text Available Main causes of accident is due to driver itself that is influenced by their bad attitude while driving. Human attitude is closely related to the human psychology. Apart from that, spiritual aspect also influence human attitude. Hence, this study carried out to improve driver safety using a new approach through psychology and spiritual factors. Objectives of this study are to identify then analyze factors of psychological and spiritual that contribute towards safety driving. A self-administered questionnaire were distributed among 256 respondents from various type of background. An analysis descriptive statistics show demographic and experience of respondents. Chi-square analysis showed only education level and traffic summon are significant to safety driving. Furthermore, correlation analysis shows psychological factors has strong linear relationship on attitude of drivers towards safety driving while spiritual factor, the perception of the spiritual and practices, both have a strong relationship to safety driving. Regression analysis demonstrates boths psychological and spiritual factors have strong evidence and significant relationship with safety driving. Thus, it can be identified that spiritual psychological factors encourage drivers to drive more safely and reduce road accidents. Therefore, this study propose useful guidelines to related agencies in order to enhance safety among drivers to be able drive safely on the road.

  8. Enhancing Safety Culture in Complex Nuclear Industry Projects

    International Nuclear Information System (INIS)

    Gotcheva, N.

    2016-01-01

    This paper presents an on-going research project “Management principles and safety culture in complex projects” (MAPS), supported by the Finnish Research Programme on Nuclear Power Plant Safety 2015-2018. The project aims at enhancing safety culture and nuclear safety by supporting high quality execution of complex projects in the nuclear industry. Safety-critical industries are facing new challenges, related to increased outsourcing and complexity in technology, work tasks and organizational structures (Milch and Laumann, 2016). In the nuclear industry, new build projects, as well as modernisation projects are temporary undertakings often carried out by networks of companies. Some companies may have little experience in the nuclear industry practices or consideration of specific national regulatory requirements. In large multinational subcontractor networks, the challenge for assuring nuclear safety arises partly from the need to ensure that safety and quality requirements are adequately understood and fulfilled by each partner. Deficient project management practices and unsatisfactory nuclear safety culture in project networks have been recognised as contributing factors to these challenges (INPO, 2010). Prior evidence indicated that many recent major projects have experienced schedule, quality and financial challenges both in the nuclear industry (STUK, 2011) and in the non-nuclear domain (Ahola et al., 2014; Brady and Davies, 2010). Since project delays and quality issues have been perceived mainly as economic problems, project management issues remain largely understudied in safety research. However, safety cannot be separated from other performance aspects if a systemic view is applied. Schedule and quality challenges may reflect deficiencies in coordination, knowledge and competence, distribution of roles and responsibilities or attitudes among the project participants. It is increasingly understood that the performance of the project network in all

  9. Use of a Web Site to Enhance Criticality Safety Training

    International Nuclear Information System (INIS)

    Huang, S T; Morman, J

    2003-01-01

    Currently, a website dedicated to enhancing communication and dissemination of criticality safety information is sponsored by the U.S. Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). This website was developed as part of the DOE response to the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 97-2, which reflected the need to make criticality safety information available to a wide audience. The website is the focal point for DOE nuclear criticality safety (NCS) activities, resources and references, including hyperlinks to other sites actively involved in the collection and dissemination of criticality safety information. The website is maintained by the Lawrence Livermore National Laboratory (LLNL) under auspices of the NCSP management. One area of the website contains a series of Nuclear Criticality Safety Engineer Training (NCSET) modules. During the past few years, many users worldwide have accessed the NCSET section of the NCSP website and have downloaded the training modules as an aid for their training programs. This trend was remarkable in that it points out a continuing need of the criticality safety community across the globe. It has long been recognized that training of criticality safety professionals is a continuing process involving both knowledge-based training and experience-based operations floor training. As more of the experienced criticality safety professionals reach retirement age, the opportunities for mentoring programs are reduced. It is essential that some method be provided to assist the training of young criticality safety professionals to replenish this limited human expert resource to support on-going and future nuclear operations. The main objective of this paper is to present the features of the NCSP website, including its mission, contents, and most importantly its use for the dissemination of training modules to the criticality safety community. We will discuss lessons learned and several ideas

  10. Replacement of the computerized control system at NPP under operation

    International Nuclear Information System (INIS)

    Ermolaev, A.D.; Rakitin, I.D.

    1985-01-01

    Reasons and preconditions for replacement of the computerized control systems (CCS) at NPP under operation are consi-- dered. Problems dealing with management of CCS replacement, maintenance of a new CCS as well as NPP personnel training for the new system maintenance are discussed. A necessity of NPP personnel participation in these works in order to adapt CCS to requirements of NPP operation personnel and to initiate the training process is underlined. Replacement of CCS at NPP under operation is associated, as a rule, with obsolescence of old systems not ensuring growing requirements to NPP workability and safety. Principles observed at CCS replacement are reduced, mainly, to the following; maximum utilizatian of existing equipment, metal strUctures, cables, instruments, power supplies, ventilation system minimum of construction works and new communications; the least change of acting panels and boxes; changes in control desks should be introduced on the basis of the analysis of operator actions '

  11. Dissemination of Knowledge about NPP Instrumentation and Control Systems

    International Nuclear Information System (INIS)

    Yastrebenetsky, M.

    2016-01-01

    Full text: Instrumentation and control (I&C) systems are the most variable part in the nuclear power plants (NPP) comparatively with any other NPP systems. This statement is connected with the wide use of computers, rapid changes in information technologies, with the appearance of new computer complex electronic components, e.g., field programmable gate arrays (FPGA) and with appropriate point of their insertion into NPP I&C life cycle. The changes in NPP I&C systems require the dissemination of the knowledge about these systems. Lessons after Fukushima accident increase necessity of these actions. The elaboration and following dissemination of this knowledge took place in different directions: • Writing and issue of three new books about NPP I&C systems for specialists and for students which were issued in Ukrainian and USA public houses (the last book was issued in 2014); • Organization of five international scientific technical conferences, devoted to NPP I&C safety problems; • Elaboration of national (Ukrainian) standards and regulations pertaining to safety important NPP I&C systems (the last standard was issued in 2015) and participation in elaboration of international standards; • Lecturing for university students, NPP specialists and I&C designers. These actions in all directions are added to IAEA activity in the area NPP I&C systems (e.g., IAEA NP-T-3.12 “Core Knowledge on I&C systems in NPP”). (author

  12. Safety parameter display system: an operator support system for enhancement of safety in Indian PHWRs

    International Nuclear Information System (INIS)

    Subramaniam, K.; Biswas, T.

    1994-01-01

    Ensuring operational safety in nuclear power plants is important as operator errors are observed to contribute significantly to the occurrence of accidents. Computerized operator support systems, which process and structure information, can help operators during both normal and transient conditions, and thereby enhance safety and aid effective response to emergency conditions. An important operator aid being developed and described in this paper, is the safety parameter display system (SPDS). The SPDS is an event-independent, symptom-based operator aid for safety monitoring. Knowledge-based systems can provide operators with an improved quality of information. An information processing model of a knowledge based operator support system (KBOSS) developed for emergency conditions using an expert system shell is also presented. The paper concludes with a discussion of the design issues involved in the use of a knowledge based systems for real time safety monitoring and fault diagnosis. (author). 8 refs., 4 figs., 1 tab

  13. Planning and management support for NPP personnel SAT-based training programmes

    International Nuclear Information System (INIS)

    Ziakova, M.

    1998-01-01

    This paper deals with planning and management support for NPP personnel SAT based training programmes based on IAEA TC Project SLR/0/003 on upgrading NPP personnel training, with the aim of upgrading NPP safety and reliability of NPP operation and maintenance. The costs needed include both Slovak and IAEA sources. Five stages of the Project are defined: planning; organizing; motivating; implementation; control, review and accountability

  14. Project No. 7 - Decommissioning unit at Ignalina NPP. (Engineering and project management)

    International Nuclear Information System (INIS)

    2000-01-01

    In order to manage decommissioning process at Ignalina NPP, a new unit should be established in the framework of the Ignalina NPP organizational structure. The on-site Engineering Project Management Unit (PMU) will adopt an integrated approach to project management , engineering design, planning, procurement, safety and licensing activities at Ignalina NPP site. The PMU will assist the Ignalina NPP management in the development of an integrated decommissioning and waste management strategy

  15. Lessons learned from current Qinshan CANDU project and the impact on future NPP's

    International Nuclear Information System (INIS)

    Hedges, K. R.; Didsbury, R.; Yu, S. K. W.

    2000-01-01

    AECL has adopted an evolutionary approach to the development of the CANDU 6 and CANDU 9 Nuclear Power Plant (NPP) designs. Each new NPP project benefits from previous projects and contains an increasing number of fully proven enhancements. In accordance with this evolutionary design approach, AECL has built on the Wolsong and Qinshan successes and the solid performance of the reference CANDU stations to define, review and implement the enhancements for the CANDU 9 NPP. Some of these enhancements include fully integrated project information systems and databases, safety enhancements coming from PSA studies and licensing activities, distributed control systems for plant-wide control and an advanced control center which addresses human factors engineering concepts. Examples of the Qinshan CANDU project delivery enhancements are the utilization of electronic engineering tools for the complete plant, and the linking of these tools with the project material management system and document management systems. The project information is reviewed and approved at the engineering office in Canada and then transmitted to site electronically. Once the electronic data is at site the information packages are extracted as necessary to enable construction and facilitate contract needs with minimum effort. This paper will provide details of the CANDU Qinshan project experiences as well as describing some of the corresponding CANDU 9 enhancements. (author)

  16. Criticality safety enhancements for SCALE 6.2 and beyond

    International Nuclear Information System (INIS)

    Rearden, Bradley T.; Bekar, Kursat B.; Celik, Cihangir; Clarno, Kevin T.; Dunn, Michael E.; Hart, Shane W.; Ibrahim, Ahmad M.; Johnson, Seth R.; Langley, Brandon R.; Lefebvre, Jordan P.; Lefebvre, Robert A.; Marshall, William J.; Mertyurek, Ugur; Mueller, Don; Peplow, Douglas E.; Perfetti, Christopher M.; Petrie Jr, Lester M.; Thompson, Adam B.; Wiarda, Dorothea; Wieselquist, William A.; Williams, Mark L.

    2015-01-01

    SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. Since 1980, regulators, industry, and research institutions around the world have relied on SCALE for nuclear safety analysis and design. SCALE 6.2 provides several new capabilities and significant improvements in many existing features for criticality safety analysis. Enhancements are realized for nuclear data; multigroup resonance self-shielding; continuous-energy Monte Carlo analysis for sensitivity/uncertainty analysis, radiation shielding, and depletion; and graphical user interfaces. An overview of these capabilities is provided in this paper, and additional details are provided in several companion papers.

  17. Safety culture enhancement through the implementation of IAEA guidelines

    International Nuclear Information System (INIS)

    Mengolini, A.; Debarberis, L.

    2007-01-01

    This paper presents the methodology applied and the results achieved in adapting and implementing the IAEA guidelines on safety culture to a research reactor as a step towards supporting its Life Management Program. The background is presented together with the effort undertaken to develop awareness on safety culture and the enhancement programme hereafter developed. The present study shows how issues of safety culture, management awareness and commitment deserve attention and can be of fundamental relevance also for research reactors. The study presents how guidelines developed specifically for nuclear power installations (NPPs) can be adapted to meet the needs and peculiarities of other nuclear installations. Moreover, the difficulties met during the implementation of the guidelines are discussed and important information and lessons can be learnt for the nuclear industry in general

  18. Operating experience: safety perspective

    International Nuclear Information System (INIS)

    Piplani, Vivek; Krishnamurthy, P.R.; Kumar, Neeraj; Upadhyay, Devendra

    2015-01-01

    Operating Experience (OE) provides valuable information for improving NPP safety. This may include events, precursors, deviations, deficiencies, problems, new insights to safety, good practices, lessons and corrective actions. As per INSAG-10, an OE program caters as a fundamental means for enhancing the defence-in-depth at NPPs and hence should be viewed as ‘Continuous Safety Performance Improvement Tool’. The ‘Convention on Nuclear Safety’ also recognizes the OE as a tool of high importance for enhancing the NPP safety and its Article 19 mandates each contracting party to establish an effective OE program at operating NPPs. The lessons drawn from major accidents at Three Mile Island, Chernobyl and Fukushima Daiichi NPPs had prompted nuclear stalwarts to change their safety perspective towards NPPs and to frame sound policies on issues like safety culture, severe accident prevention and mitigation. An effective OE program, besides correcting current/potential problems, help in proactively improving the NPP design, operating and maintenance procedures, practices, training, etc., and thus plays vital role in ensuring safe and efficient operation of NPPs. Further it enhances knowledge with regard to equipment operating characteristics, system performance trends and provides data for quantitative and qualitative safety analysis. Besides all above, an OE program inculcates a learning culture in the organisation and thus helps in continuously enhancing the expertise, technical competency and knowledge base of its staff. Nuclear and Radiation Facilities in India are regulated by Atomic Energy Regulatory Board (AERB). Operating Plants Safety Division (OPSD) of AERB is involved in managing operating experience activities. This paper provides insights about the operating experience program of OPSD, AERB (including its on-line data base namely OPSD STAR) and its utilisation in improving the regulations and safety at Indian NPPs/projects. (author)

  19. Enhanced Maritime Safety through Diagnosis and Fault Tolerant Control

    DEFF Research Database (Denmark)

    Blanke, Mogens

    2001-01-01

    Faults in steering, navigation instruments or propulsion machinery are serious on a marine vessel since the consequence could be loss of maneuvering ability, and imply risk of damage to vessel personnel or environment. Early diagnosis and accomodation of faults could enhance safety. Fault...... of properties of a falty system; means to determine remedial actions. The paper illustrates the techniques by two marine examples, sensor fusion for automatic steering and control of the main engine....

  20. Perspective channel-type reactor with enhanced safety

    International Nuclear Information System (INIS)

    Adamov, E.O.; Grozdov, I.I.; Kuznetsov, S.P.; Petrov, A.A.; Rozhdestvensky, M.I.; Cherkashov, Yu.M.

    1994-01-01

    Following the search for new design solutions to develop within the framework of channel trends the reactor with enhanced safety the Research and Development Institute of Power Engineering has developed the design of the multiloop boiling water reactor (MKER). The MKER enhanced safety is attained when involving the inherent safety features, passive safety systems as well as the accident consequences confinement devices. The design realizes several advantages which are typical of the channel-type reactors, namely: The design desintegration simplifying the manufacture, control, equipment delivery and decreasing, versus the pressure vessel reactors, the accident effect if it proceeds in an explosive manner; small operating reactivity margin and fuel burnup increased due to continuous refuelling; fuel cycle flexibility allowing comparatively easily to adopt the reactor to the conjuncture of the country fuel balance; multiloop circuit of the main coolant which reduces the degree and effect of the accidents connected with the equipment and pipings rupture; monitoring of the channels and fuel assemblies leak-tightness. (orig.)

  1. Design verification enhancement of field programmable gate array-based safety-critical I&C system of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Ibrahim [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 17104 (Korea, Republic of); Jung, Jaecheon, E-mail: jcjung@kings.ac.kr [Department of Nuclear Power Plant Engineering, KEPCO International Nuclear Graduate School, 658-91 Haemaji-ro, Seosang-myeon, Ulju-gun, Ulsan 45014 (Korea, Republic of); Heo, Gyunyoung [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 17104 (Korea, Republic of)

    2017-06-15

    Highlights: • An enhanced, systematic and integrated design verification approach is proposed for V&V of FPGA-based I&C system of NPP. • RPS bistable fixed setpoint trip algorithm is designed, analyzed, verified and discussed using the proposed approaches. • The application of integrated verification approach simultaneously verified the entire design modules. • The applicability of the proposed V&V facilitated the design verification processes. - Abstract: Safety-critical instrumentation and control (I&C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. However, safety analysis for FPGA-based I&C systems, and verification and validation (V&V) assessments still remain important issues to be resolved, which are now become a global research point of interests. In this work, we proposed a systematic design and verification strategies from start to ready-to-use in form of model-based approaches for FPGA-based reactor protection system (RPS) that can lead to the enhancement of the design verification and validation processes. The proposed methodology stages are requirement analysis, enhanced functional flow block diagram (EFFBD) models, finite state machine with data path (FSMD) models, hardware description language (HDL) code development, and design verifications. The design verification stage includes unit test – Very high speed integrated circuit Hardware Description Language (VHDL) test and modified condition decision coverage (MC/DC) test, module test – MATLAB/Simulink Co-simulation test, and integration test – FPGA hardware test beds. To prove the adequacy of the proposed

  2. Design verification enhancement of field programmable gate array-based safety-critical I&C system of nuclear power plant

    International Nuclear Information System (INIS)

    Ahmed, Ibrahim; Jung, Jaecheon; Heo, Gyunyoung

    2017-01-01

    Highlights: • An enhanced, systematic and integrated design verification approach is proposed for V&V of FPGA-based I&C system of NPP. • RPS bistable fixed setpoint trip algorithm is designed, analyzed, verified and discussed using the proposed approaches. • The application of integrated verification approach simultaneously verified the entire design modules. • The applicability of the proposed V&V facilitated the design verification processes. - Abstract: Safety-critical instrumentation and control (I&C) system in nuclear power plant (NPP) implemented on programmable logic controllers (PLCs) plays a vital role in safe operation of the plant. The challenges such as fast obsolescence, the vulnerability to cyber-attack, and other related issues of software systems have currently led to the consideration of field programmable gate arrays (FPGAs) as an alternative to PLCs because of their advantages and hardware related benefits. However, safety analysis for FPGA-based I&C systems, and verification and validation (V&V) assessments still remain important issues to be resolved, which are now become a global research point of interests. In this work, we proposed a systematic design and verification strategies from start to ready-to-use in form of model-based approaches for FPGA-based reactor protection system (RPS) that can lead to the enhancement of the design verification and validation processes. The proposed methodology stages are requirement analysis, enhanced functional flow block diagram (EFFBD) models, finite state machine with data path (FSMD) models, hardware description language (HDL) code development, and design verifications. The design verification stage includes unit test – Very high speed integrated circuit Hardware Description Language (VHDL) test and modified condition decision coverage (MC/DC) test, module test – MATLAB/Simulink Co-simulation test, and integration test – FPGA hardware test beds. To prove the adequacy of the proposed

  3. Boron analyses in the reactor coolant system of French PWR by acid-base titration ([B]) and ICP-MS (10B atomic %): key to NPP safety

    International Nuclear Information System (INIS)

    Jouvet, Fabien; Roux, Sylvie; Carabasse, Stephanie; Felgines, Didier

    2012-09-01

    Boron is widely used by Nuclear Power Plants and especially by EDF Pressurized Water Reactors to ensure the control of the neutron rate in the reactor coolant system and, by this way, the fission reaction. The Boron analysis is thus a major factor of safety which enables operators to guarantee the permanent control of the reactor. Two kinds of analyses carried out by EDF on the Boron species, recently upgraded regarding new method validation standards and developed to enhance the measurement quality by reducing uncertainties, will be discussed in this topic: Acid-Base titration of Boron and Boron isotopic composition by Inductively Coupled Plasma Mass Spectrometer - ICP MS. (authors)

  4. Accident management insights after the Fukushima Daiichi NPP accident

    International Nuclear Information System (INIS)

    Degueldre, Didier; Viktorov, Alexandre; Tuomainen, Minna; Ducamp, Francois; Chevalier, Sophie; Guigueno, Yves; Tasset, Daniel; Heinrich, Marcus; Schneider, Matthias; Funahashi, Toshihiro; Hotta, Akitoshi; Kajimoto, Mitsuhiro; Chung, Dae-Wook; Kuriene, Laima; Kozlova, Nadezhda; Zivko, Tomi; Aleza, Santiago; Jones, John; McHale, Jack; Nieh, Ho; Pascal, Ghislain; ); Nakoski, John; Neretin, Victor; Nezuka, Takayoshi; )

    2014-01-01

    The Fukushima Daiichi nuclear power plant (NPP) accident, that took place on 11 March 2011, initiated a significant number of activities at the national and international levels to reassess the safety of existing NPPs, evaluate the sufficiency of technical means and administrative measures available for emergency response, and develop recommendations for increasing the robustness of NPPs to withstand extreme external events and beyond design basis accidents. The OECD Nuclear Energy Agency (NEA) is working closely with its member and partner countries to examine the causes of the accident and to identify lessons learnt with a view to the appropriate follow-up actions to be taken by the nuclear safety community. Accident management is a priority area of work for the NEA to address lessons being learnt from the accident at the Fukushima Daiichi NPP following the recommendations of Committee on Nuclear Regulatory Activities (CNRA), Committee on the Safety of Nuclear Installations (CSNI), and Committee on Radiation Protection and Public Health (CRPPH). Considering the importance of these issues, the CNRA authorised the formation of a task group on accident management (TGAM) in June 2012 to review the regulatory framework for accident management following the Fukushima Daiichi NPP accident. The task group was requested to assess the NEA member countries needs and challenges in light of the accident from a regulatory point of view. The general objectives of the TGAM review were to consider: - enhancements of on-site accident management procedures and guidelines based on lessons learnt from the Fukushima Daiichi NPP accident; - decision-making and guiding principles in emergency situations; - guidance for instrumentation, equipment and supplies for addressing long-term aspects of accident management; - guidance and implementation when taking extreme measures for accident management. The report is built on the existing bases for capabilities to respond to design basis

  5. Radioactive source management in Daya Bay NPP

    International Nuclear Information System (INIS)

    Mao Chun Yang

    2000-01-01

    'Small sources causes big accidents' had occurred worldwide many times. Radioactive source management in Nuclear Power Plant in very important for its safety record. This paper introduces the way and experience of radioactive source management in Daya Bay NPP from aspects of clarifying the responsibilities, centralizing the management of high radioactivity sources, work process management and experience feedback etc. (author)

  6. External hazards considered for Paks NPP

    International Nuclear Information System (INIS)

    Kiss, Tibor

    2000-01-01

    PAKS NPP was built according to Soviet construction standards which took into account meteorological aspects but no documents for other external hazards were available. Main activities concerning earthquakes cover reevaluation of the plant site, seismic safety technological concept, improving the seismic resistance, installation of seismic monitoring and protection system, and seismic PSA

  7. Slovakia: Mochovce NPP. Project control. Annex 10

    International Nuclear Information System (INIS)

    Bodorik, K.

    1999-01-01

    This annex deals with project control. Mochovce NPP suffered considerable delay primarily due to lack of money. This situation was corrected and construction resumed in 1996. Throughout the 'dormant' period the plant received considerable support from the major contractors, who maintained skeleton staff at site. Significant safety and managerial improvements are being introduced and a strategic plan for the plant has been developed. (author)

  8. 78 FR 76391 - Proposed Enhancements to the Motor Carrier Safety Measurement System (SMS) Public Web Site

    Science.gov (United States)

    2013-12-17

    ...-0392] Proposed Enhancements to the Motor Carrier Safety Measurement System (SMS) Public Web Site AGENCY... proposed enhancements to the display of information on the Agency's Safety Measurement System (SMS) public Web site. On December 6, 2013, Advocates [[Page 76392

  9. Reactor plant for Belene NPP completion

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Ryzhov, S. B.; Ermakov, D. N.; Repin, A. I.

    2004-01-01

    Construction of 'Belene' NPP was started at the end of 80-ties using project U-87 with V-320 reactor plant, general designer of this plant is OKB 'Gidropress'. At the beginning of 90-ties, on completing the considerable number of deliveries and performance of civil engineering work at the site the NPP construction was suspended. Nowadays, considering the state of affairs at the site and the work performed by Bulgarian Party on preservation of the equipment delivered, the most perspective is supposed to be implementation of the following versions in completing 'Belene' NPP: for completion of Unit 1 - reactor plant VVER-1000 on the basis of V-320 reactor with the maximum use of the delivered equipment (V-320M) having the extended service life and safety improvement; for Unit 2 - advanced reactor plant VVER-1000. For the upgraded reactor plant V-230M the basic solutions and characteristics are presented, as well as the calculated justification of strength and safety analyses, design of the reactor core and fuel cycle, instrumentation and control systems, application of the 'leak-before break' in the project and implementation of safety measures. For the modernised reactor plant V-392M the main characteristics and basic changes are presented, concerning reactor pressure vessel, steam generator, reactor coolant pump set. Design of NPP with the modernized reactor plant V-320M meets the up-to-date requirements and can be licensed for completion and operation. In the design of NPP with the advanced reactor plant the basic solutions and the equipment are used that are similar to those used in standard reactor plant V-320 and new one with VVER-1000 under construction and completion in Russia, and abroad. Compliance of reactor design with the up-to-date international requirements, considering the extended service life of the main equipment, shows its rather high potential for implementation during completion of 'Belene' NPP

  10. Public opinion concerning completion of Khmelnitsky-2 and Rovno-4 NPP's construction for compensation of Chernobyl NPP shut-down

    International Nuclear Information System (INIS)

    Litvins'kij, L.L.; Novos'olov, G.M.; Majstrenko, O.V.; Sazonov, G.V.; Gryin, L.

    2001-01-01

    Interrogation of representation from non-government organizations and private persons from 46 countries about their opinion on completion of two NPP's construction revealed many comments about environmental effects, safety and economic aspects. These comments are analyzed by specialists

  11. Real-time management (RTM) by cloud computing system dynamics (CCSD) for risk analysis of Fukushima nuclear power plant (NPP) accident

    International Nuclear Information System (INIS)

    Cho, Hyo Sung; Woo, Tae Ho; The Cyber Univ. of Korea, Seoul

    2017-01-01

    The earthquake and tsunami induced accident of nuclear power plant (NPP) in Fukushima disaster is investigated by the real-time management (RTM) method. This non-linear logic of the safety management is applied to enhance the methodological confidence in the NPP reliability. The case study of the earthquake is modeled for the fast reaction characteristics of the RTM. The system dynamics (SD) modeling simulations and cloud computing are applied for the RTM method where the real time simulation has the fast and effective communication for the accident remediation and prevention. Current tablet computing system can improve the safety standard of the NPP. Finally, the procedure of the cloud computing system dynamics (CCSD) modeling is constructed.

  12. Real-time management (RTM) by cloud computing system dynamics (CCSD) for risk analysis of Fukushima nuclear power plant (NPP) accident

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hyo Sung [Yonsei Univ., Wonju Gangwon-do (Korea, Republic of). Dept. of Radiation Convergence Engineering; Woo, Tae Ho [Yonsei Univ., Wonju Gangwon-do (Korea, Republic of). Dept. of Radiation Convergence Engineering; The Cyber Univ. of Korea, Seoul (Korea, Republic of). Dept. of Mechanical and Control Engineering

    2017-03-15

    The earthquake and tsunami induced accident of nuclear power plant (NPP) in Fukushima disaster is investigated by the real-time management (RTM) method. This non-linear logic of the safety management is applied to enhance the methodological confidence in the NPP reliability. The case study of the earthquake is modeled for the fast reaction characteristics of the RTM. The system dynamics (SD) modeling simulations and cloud computing are applied for the RTM method where the real time simulation has the fast and effective communication for the accident remediation and prevention. Current tablet computing system can improve the safety standard of the NPP. Finally, the procedure of the cloud computing system dynamics (CCSD) modeling is constructed.

  13. NPP Krsko small break LOCA analysis

    International Nuclear Information System (INIS)

    Mavko, B.; Petelin, S.; Peterlin, G.

    1987-01-01

    Parametric analysis of small break loss of coolant accident for the Krsko NPP was calculated by using RELAP5/MOD1 computer code. The model that was used in our calculations has been improved over several years and was previously tested in simulation (s) of start-up tests and known NPP Krsko transients. In our calculations we modelled automatic actions initiated by control, safety and protection systems. We also modelled the required operator actions as specified in emergency operating instructions. In small-break LOCA calculations, we varied break sizes in the cold leg. The influence of steam generator tube plugging on small break LOCA accidents was also analysed. (author)

  14. Nicotine disrupts safety learning by enhancing fear associated with a safety cue via the dorsal hippocampus.

    Science.gov (United States)

    Connor, David A; Kutlu, Munir G; Gould, Thomas J

    2017-07-01

    Learned safety, a learning process in which a cue becomes associated with the absence of threat, is disrupted in individuals with post-traumatic stress disorder (PTSD). A bi-directional relationship exists between smoking and PTSD and one potential explanation is that nicotine-associated changes in cognition facilitate PTSD emotional dysregulation by disrupting safety associations. Therefore, we investigated whether nicotine would disrupt learned safety by enhancing fear associated with a safety cue. In the present study, C57BL/6 mice were administered acute or chronic nicotine and trained over three days in a differential backward trace conditioning paradigm consisting of five trials of a forward conditioned stimulus (CS)+ (Light) co-terminating with a footshock unconditioned stimulus followed by a backward CS- (Tone) presented 20 s after cessation of the unconditioned stimulus. Summation testing found that acute nicotine disrupted learned safety, but chronic nicotine had no effect. Another group of animals administered acute nicotine showed fear when presented with the backward CS (Light) alone, indicating the formation of a maladaptive fear association with the backward CS. Finally, we investigated the brain regions involved by administering nicotine directly into the dorsal hippocampus, ventral hippocampus, and prelimbic cortex. Infusion of nicotine into the dorsal hippocampus disrupted safety learning.

  15. NPP service life management

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2001-01-01

    Problems of NPP service life management and service life prolongation are reviewed. Methods for the prolongation of the French NPP service life are discussed, priority directions of nuclear block service life management in regard to aging in the context of the European program of investigation into the materials aging are identified. Questions of the provision of the 60 years service life of the Mihama 1 block (Japan) and decision of the problem of the control equipment aging in Great Britain are discussed. Situation with the prolongation of licenses on the NPP operation in the USA and Spain is considered [ru

  16. Enhancement of safety at nuclear facilities in Pakistan

    International Nuclear Information System (INIS)

    Ahmad, S.A.; Hayat, T.; Azhar, W.

    2006-01-01

    Pakistan is benefiting from nuclear technology mostly in health and energy sectors as well as agriculture and industry and has an impeccable safety record. At the national level uses of nuclear technology started in 1955 resulting in the operation of Karachi Radioisotope Center, Karachi, in December 1960. Pakistan Nuclear Safety Committee (PNSC) was formulated in 1964 with subsequent promulgation of Pakistan Atomic Energy Commission (PAEC) Ordinance in 1965 to cope with the anticipated introduction of a research reactor, namely PARR-I, and a nuclear power plant, namely KANUPP. Since then Pakistan's nuclear program has expanded to include numerous nuclear facilities of varied nature. This program has definite economic and social impacts by producing electricity, treating and diagnosing cancer patients, and introducing better crop varieties. Appropriate radiation protection includes a number of measures including database of sealed radiation sources at PAEC operated nuclear facilities, see Table l, updated during periodic physical verification of these sources, strict adherence to the BSS-115, IAEA recommended enforcement of zoning at research reactors and NPPs, etc. Pakistan is party to several international conventions and treaties, such as Convention of Nuclear Safety and Early Notification, to improve and enhance safety at its nuclear facilities. In addition Pakistan generally and PAEC particularly believes in a blend of prudent regulations and good/best practices. This is described in this paper. (Author)

  17. Account of requirements for modernization in VPBER-600 enhanced safety reactor instrumentation and control system development

    International Nuclear Information System (INIS)

    Shashkin, S.L.; Pobedonostsev, A.B.; Drumov, V.V.; Chudin, A.G.

    1993-01-01

    Nuclear power plant (NPP) with VPBER-600 reactor is a station of new generation. The specified term of reactor plant operation is 60 years and taking into account that the proposed term of starting the first power unit is on the turn of centuries one can definitely state that for Russia conditions VPBER-600 is a plant of 21 century. Such far removed term for NPP now in the stage of development as it can seem does not put the problems of modernization as first order tasks. But open-quotes...who does not think about future lives in the past.close quotes It is that the NPP instrumentation and control (I ampersand C) systems are in the most degree subjected to the influence of factors which favor their modifications. These factors can be arbitrarily divided into two groups: (1) inner factors, i.e. changes (failures, aging, etc) in I ampersand C components as well as changes dictated by technological reasons (change of equipment composition, control algorithms, operation modes); (2) outer factors, i.e. intensive development of information technologies and rapid improvement of electronic components. This presentation addresses the problem of modernization of the safety instrumentation for this next generation facility, and the research effort it will entail. The system is designed to allow for modernization, and the relatively easy adoption of new instrumentation and technology as it becomes available

  18. Seismic response analysis and upgrading design of pump houses of Kozloduy NPP units 5 and 6

    International Nuclear Information System (INIS)

    Jordanov, M.; Marinov, M.; Krutzik, N.

    2001-01-01

    The main objective of the presented project was to perform a feasibility study for seismic/structural evaluation of the safety related structures at Kozloduy NPP Units 5 and 6 for the new site seismicity and determine if they satisfy current international safety standards. The evaluation of the Pump House 3 (PH3) building is addressed in this paper, which was carried out by applying appropriate modeling techniques combined with failure mode and seismic margin analyses. The scope of the work defined was to present the required enhancement of the seismic capacity of the Pump House structures.(author)

  19. Study on new-type fuel-related assembly handling tools for PWR NPP

    International Nuclear Information System (INIS)

    Fan Xiumei

    2013-01-01

    This article describes the design and study on a set of new-type fuel-related assembly snatching tools used for PWR NPP. The purpose is mainly to enhance the tool safety, reliability and convenientness by improvement of the mechanism and structure of the tool for snatching preciseness and avoiding from falling and abrasion of fuel-related assemblies for any condition. The new-type fuel-related assembly handling tools are compared with similar equipment in worldwide in terms of function, main technical characteristic, and safety and protection, some of them are better than the similar equipment in that they have reliable loading and unloading and conveying capabilities. (author)

  20. Effective corrective actions to enhance operational safety of nuclear installations

    International Nuclear Information System (INIS)

    2005-07-01

    The safe operation of nuclear power plants around the world and the prevention of incidents in these installations remain key concerns for the nuclear community. In this connection the feedback of operating experience plays a major role: every nuclear plant operator needs to have a system in place to identify and feed back the lessons learned from operating experience and to implement effective corrective actions to prevent safety events from reoccurring. An effective operating experience programme also includes a proactive approach that is aimed at preventing the first-time occurrence of safety events. In April 2003, the IAEA issued the PROSPER guidelines for nuclear installations to strengthen and enhance their own operating experience process and for self-assessment on the effectiveness of the feedback process. Subsequently, in the course of the Operational Safety Review Teams missions conducted by the IAEA that focused on the operational safety practices of nuclear power plants, the IAEA enhanced the review of the operating experience in nuclear power plants by implementing a new module that is derived from these guidelines. In order to highlight the effective implementation of the operating experience programme and to provide practical assistance in this area, the IAEA organized workshops and conferences to discuss recent trends in operating experience. The IAEA also performed assistance and review missions at plants and corporate organizations. The IAEA is further developing advice and assistance on operating experience feedback programmes and is reporting on good practices. The present publication is the outcome of two years of coordinated effort involving the participation of experts of nuclear organizations in several Member States. It provides information and good practices for successfully establishing an effective corrective actions programme. This publication forms part of a series that develops the principles set forth in these guidelines

  1. National plan to enhance aviation safety through human factors improvements

    Science.gov (United States)

    Foushee, Clay

    1990-01-01

    The purpose of this section of the plan is to establish a development and implementation strategy plan for improving safety and efficiency in the Air Traffic Control (ATC) system. These improvements will be achieved through the proper applications of human factors considerations to the present and future systems. The program will have four basic goals: (1) prepare for the future system through proper hiring and training; (2) develop a controller work station team concept (managing human errors); (3) understand and address the human factors implications of negative system results; and (4) define the proper division of responsibilities and interactions between the human and the machine in ATC systems. This plan addresses six program elements which together address the overall purpose. The six program elements are: (1) determine principles of human-centered automation that will enhance aviation safety and the efficiency of the air traffic controller; (2) provide new and/or enhanced methods and techniques to measure, assess, and improve human performance in the ATC environment; (3) determine system needs and methods for information transfer between and within controller teams and between controller teams and the cockpit; (4) determine how new controller work station technology can optimally be applied and integrated to enhance safety and efficiency; (5) assess training needs and develop improved techniques and strategies for selection, training, and evaluation of controllers; and (6) develop standards, methods, and procedures for the certification and validation of human engineering in the design, testing, and implementation of any hardware or software system element which affects information flow to or from the human.

  2. EC6 safety enhancement - including impact of Fukushima lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Yu, S.; Zemdegs, R.; Boyle, S.; Soulard, M., E-mail: stephen.yu@candu.com [Candu Energy Inc., Mississauga, Ontario (Canada)

    2012-09-15

    The Enhanced CANDU 6 (EC6) is the new Generation III CANDU reactor design that meets the most up to date regulatory requirements and customer expectations. EC6 builds on the proven high performance design inch as the Qinshan CANDU 6 units and has made improvements to safety and operational performance, and has incorporated extensive operational feedback including Fukushima. The Fukushima Dai-ichi March 11, 2011 event has demonstrated the importance of defence-in-depth considerations for beyond-design basis events, including severe accidents. The EC6 design is based on the defence-in-depth principles and provides further design features that address the lessons learned from Fukushima. (author)

  3. Safeguards at Kozloduy NPP - Experience and expectations

    International Nuclear Information System (INIS)

    Elenkov, Todor

    2001-01-01

    Bulgaria is a party of Non Proliferation Treaty since 5 September 1969. The agreement between IAEA and Bulgaria - INFCIRC 178 - has been in force since 29 February 1972. At that time Bulgaria had one research reactor IRT-2000 in Sofia and two power reactors of WWER-440 type under construction. Now at Kozloduy NPP site there are 4 facilities, which consist of 4 WWER-440 and 2 WWER-1000 type power reactors, producing almost 50% of the electricity in Bulgaria and 1 wet away from reactor spent fuel storage. In 1991 under the green movements and social pressure, the research reactor in Sofia was closed and the construction of the second NPP in Belene with 2 WWER-1000 type reactors was halted. After the transfer in 1994 of the fresh fuel from the research reactor to Kozloduy due to security reasons practically NPP Kozloduy remains the only significant (from safeguards point of view) nuclear site in Bulgaria. In 1972 a 'Nuclear Fuel' group was formed at the Physicists Department in NPP Kozloduy with responsibilities to carry out for safeguards records and reports, fresh and spent fuel transport and control. In 1990 this group was transferred to the Safety Section and since 1992 it exists as 'Control and Accounting for of the Nuclear Materials' - a section in the Safety Department. Currently the section serves all four facilities in NPP Kozloduy and has four people: section head, chief inspector and two inspectors. The main activities of the section include: a) Control of the nuclear fuel location as well as meeting the storage and transport conditions regulations; b) Control of the conditions for normal operations of the installed IAEA surveillance systems; c) Preparation of documents for licensing of fresh and spent nuclear fuel transport; d) Preparation of the official information on nuclear materials location and quantity; e) Preparation of accounting records and the reports for IAEA (ICR, PIL, MBR); f) Co-ordination of the IAEA safeguards inspection activities at NPP

  4. Management of safety culture

    International Nuclear Information System (INIS)

    Kavsek, D.

    2004-01-01

    The strengthening of safety culture in an organization has become an increasingly important issue for nuclear industry. A high level of safety performance is essential for business success in intensely competitive global environment. This presentation offers a discussion of some principles and activities used in enhancing safety performance and appropriate safety behaviour at the Krsko NPP. Over the years a number of events have occurred in nuclear industry that have involved problems in human performance. A review of these and other significant events has identified recurring weaknesses in plant safety culture and policy. Focusing attention on the strengthening of relevant processes can help plants avoid similar undesirable events. The policy of the Krsko NPP is that all employees concerned shall constantly be alert to opportunities to reduce risks to the lowest practicable level and to achieve excellence in plant safety. The most important objective is to protect individuals, society and the environment by establishing and maintaining an effective defense against radiological hazard in the nuclear power plant. It is achieved through the use of reliable structures, components, systems, and procedures, as well as plant personnel committed to a strong safety culture. The elements of safety culture include both organizational and individual aspects. Elements commonly included at the organizational level are senior management commitment to safety, organizational effectiveness, effective communication, organizational learning, and a culture that encourages identification and resolution of safety issues. Elements identified at the individual level include personal accountability, a questioning attitude, communication, procedural adherence, etc.(author)

  5. Level 1 shutdown and low power operation of Mochovce NPP, Unit 1, Slovakia

    International Nuclear Information System (INIS)

    Halada, P.; Cillik, I.; Stojka, T.; Kuzma, M.; Prochaska, J.; Vrtik, L.

    2004-01-01

    The paper presents general approach, used methods and form of documentation of the results that have been applied within the shutdown and low power PSA (SPSA) study for Mochovce NPP, Unit 1, Slovakia. The SPSA project was realized by VUJE Trnava Inc., Slovakia in 2001-2002 years. The Level 1 SPSA study for Mochovce NPP Unit 1 covers internal events as well as internal (fires, floods and heavy load drop) and external (aircraft crash, extreme meteorological conditions, seismic event and influence of surrounding industry) hazards. Mochovce NPP consists of two operating units equipped with VVER 440/V213 reactors safety upgraded before construction finishing and operation start. 87 safety measures based on VVER 440 operational experience and international mission insights were implemented to enhance its operational and nuclear safety. The SPSA relates to full power PSA (FPSA) as a continuation of the effort to create a harmonized level 1 PSA model for all operational modes of the plant with the goal to use it for further purposes as follows: Real Time Risk Monitor, Maintenance Optimization, Technical Specifications Optimization, Living PSA. (author)

  6. A study on the Methodology for Integrated Safety Assessment for Accidental Analysis on LILW managed in Temporary Storage Facility (TSF) at NPP

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Min Ho; Lee, Kun Jai [Korea Advanced Institute of Science and Technology, 335 Gwahak-ro, yuseong-gu, Daejeon 305-701 (Korea, Republic of); Choi, Kyung Woo [Korea Institute of Nuclear Safety, 34 Gwahak-ro, yuseong-gu, Daejeon 305-338 (Korea, Republic of)

    2010-07-01

    In Korea, the low and intermediate level wastes (LILW) generated from the operation of NPPs have been gradually increased since 1978 because of the absence of the repository for the disposal of LILW. And then, the LILW has being stored in the temporary storage facility (TSF) at each reactor sites. Fortunately, in 2005, the Gyeongju city was selected by the resident's vote in four competitive provinces for the LILW disposal facility. The repository for the disposal of LILW will be operated in end of 2009. In opposition to many of researches on the disposal of LILW, however, the risk assessment on the TSF has scarcely been conducted. Furthermore, the details in regards of the safety analysis on this facility have not been considered in the preliminary and final safety analysis report because this report focused on the nuclear reactor system rather than this facility. As a consequence of these situations, the number of the researches on the arbitrary accidents occurring in the TSF has not been enough. The objective of this study is to establish arbitrary accident scenarios and to evaluate the exposure dose in terms of the effective dose and the thyroid equivalent dose due to the arbitrary accidents originating in the TSF for LILW management. For the establishment of arbitrary accident scenarios in the TSF for LILW, the initiating event was derived by master logic diagram (MLD) method based on the fault tree analysis (FTA), and then arbitrary accident scenarios were developed by the event tree analysis (ETA) through the derived initiating events. The main initiating events led to the arbitrary operational accident, which is the dropping of a drum and fire, were derived from MLD method. The exposure effects resulted from release of radioactive materials related to arbitrary accidents in the TSF for LILW can be divided with the internal exposure caused by breathing and external exposure by radioactive plume. The 13 radionuclides for internal and external exposure

  7. Accident localization system with jet condensers for VVER 440-V 230 NPP at Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Murani, J.

    1995-01-01

    The operational safety of the V1 nuclear power plant (NPP) is unsatisfactory and does not correspond to present requirements as to nuclear safety. Further NPP operation after 1995 is conditional on nuclear safety enhancement to a level comparable with that in West European countries. This aim should be achieved by a principal reconstruction involving in addition to others also backfitting the V1 NPP with technical facilities aimed at coping with a design basis accident (DBA).To cope with such an accident the Power Equipment Research Institute (VUEZ) designed an accident localization system with jet condensers. This system consists of (a) an air trap (one for each unit, mutually interconnected) with an expansion bell enclosed within, placed on a plate with 200 pipes of jet condensers passing through, and (b) a connecting duct between the hermetic zone and the air trap. The vertical jet condenser is an essential element of the system designed for steam condensation. Apart from condensation it serves as a water seal separating units 1 and 2.Demonstration tests of the jet condenser (model 1:1) condensing function were carried out at the testing unit of the All-Union Research Institute for NPP Operation (VNIIAES), Moscow in Kashir, 11-22 September 1992. These experiments proved the jet condenser ability to ensure complete condensation of the steam produced. Experimental verification of the sealing function (model 1:1) was carried out at the testing unit of the VUEZ Tlmace. These experiments concerning the dynamics and overpressure in the free space above the pool were close to the conditions in the air trap during DBA. The jet condenser height was proved to be sufficient to ensure the sealing function. Design and experimental work has been implemented in close cooperation with Russian experts Mr. V.N. Bulynin from the VNIIAES, Moscow, and Mr. M.V. Kuznecov from the Scientific and Engineering Center for Nuclear and Radiological Safety, Moscow. (orig.)

  8. Enhancement of Nuclear Safety in Korea: A Regulatory Perspective

    International Nuclear Information System (INIS)

    Chung, K.Y.

    2016-01-01

    In the aftermath of Fukushima Daiichi accident in 2011 Korean regulatory body immediately performed special inspections on nuclear power plants (NPPs) and a research reactor in Korea, and issued an enforcement order for the licensees to implement fifty Fukushima action items to address the safety issues identified by the inspections. Subsequently, the licensees have established the implementation plans for resolution of the action items. By the implementation of the action items, the possibility of severe accident due to the extreme hazards has been greatly reduced and the capabilities to mitigate the severe accident, should it occur, have been upgraded. To improve the consistency and predictability of the regulation on severe accidents, Nuclear Safety and Security Commission (NSSC) the regulatory body in Korea, is revising the regulatory framework for severe accidents. The new framework will require the licensee to enhance the capabilities for prevention and mitigation of severe accidents in view of the defence in depth principle, to assess the radiological effects from the severe accidents, and to improve current accident management procedures and guidelines necessary for the prevention and mitigation of severe accidents. This rulemaking also considers the safety principles provided by the IAEA Vienna Declaration in 2015, which require new NPPs to prevent large radioactive releases. (author)

  9. Enhancement of pressurizer safety valve operability by seating design improvement

    International Nuclear Information System (INIS)

    Moisidis, N.T.; Ratiu, M.D.

    1995-01-01

    Operating conditions specific to pressurizer safety valves (PSVs) have led to numerous problems and have caused industry and NRC concerns regarding the adequacy of spring-loaded self-actuated safety valves for reactor coolant system (RCS) overpressure protection. Specific concerns are: setpoint drift, spurious actuations, and pressure protection. Specific concerns are: setpoint drift, spurious actuations, and leakage. Based on testing and valve construction analysis of a Crosby model 6M6 PSV (Moisidis and Ratiu, 1992), it was established that the primary contributor to the valve problems is a susceptibility to weak seating. To eliminate spring instability, a new spring washer was designed, which guides the spring and precludes its rotation from the reference installed position. Results of tests performed on a prototype PSV equipped with the modified upper spring washer has shown significant improvements in valve operability and a consistent setpoint reproducibility to less than ±1% of the PSV setpoint (testing of baseline, unmodified valve, resulted in a setpoint drift of ± 2%). Enhanced valve operability will result in a significant decrease in operating and maintenance costs associated with valve maintenance and testing. In addition, the enhanced setpoint reproducibility will allow the development of a nitrogen to steam correlation for future in-house PSV testing which will result in further reductions in costs associated with valve testing

  10. Enhancement of pressurizer safety valve operability by seating design improvement

    International Nuclear Information System (INIS)

    Moisidis, N.T.; Ratiu, M.D.

    1994-01-01

    Operating conditions specific to Pressurizer Safety Valves (PSVs) have led to numerous problems and have caused industry and NRC concerns regarding the adequacy of spring loaded self-actuated safety valves for Reactor Coolant System (RCS) overpressure protection. Specific concerns are: setpoint drift, spurious actuations and leakage. Based on testing and valve construction analysis of a Crosby model 6M6 PSV, it was established that the primary contributor to the valve problems is a susceptibility to weak seating. To eliminate spring instability, a new spring washer was designed, which guides the spring and precludes its rotation from the reference installed position. Results of tests performed on a prototype PSV equipped with the modified upper spring washer has shown significant improvements in valve operability and a consistent setpoint reproducibility to less than ±1% of the PSV setpoint (testing of baseline, unmodified valve, resulted in a setpoint drift of ±2%). Enhanced valve operability will result in a significant decrease in operating and maintenance costs associated with valve maintenance and testing. In addition, the enhanced setpoint reproducibility will allow the development of a nitrogen to steam correlation for future in-house PSV testing which will result in further reductions in costs associated with valve testing

  11. Modernization of the oldest Swedish NPP

    International Nuclear Information System (INIS)

    Hagberth, Ronald

    1998-01-01

    OKG operates three BWR units of ABB design: Oskarshamn 1 with a net capacity of 440' MW, Oskarshamn 2 of 600 MW and Oskarshamn 3 of 1160 MW. Oskarshamn 1 NPP was commissioned in 1972 as the first commercial nuclear unit in Sweden. After more than twenty years of successful operation, the unit is now also the first reactor in Sweden to undergo a large safety modernization program. In the year 2000 the Oskarshamn 1 NPP will be modernized to a high level of safety standard and ready for operation for another period of at least 20 years. Experience gained can be used when modernizing other NPPs. The investment program for life extension is reasonable and shows that NPPs can be operated with an expected life span of more than 40 years at an ever-increasing safety level and still be very competitive in a deregulated market. (author)

  12. Intelligent system for accident identification in NPP

    International Nuclear Information System (INIS)

    Hernandez, J.L.

    1998-01-01

    Accidental situations in NPP are great concern for operators, the facility, regulatory bodies and the environmental. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making initiator events with higher relative contribution to the reactor core damage occur. The intelligent System uses the results of the pre-operational Probabilistic safety Assessment and the Thermal hydraulic Safety Analysis of the NPP Juragua as source for building its knowledge base. The nucleus of the system is presented as a design of an intelligent hybrid from the combination of the artificial intelligence techniques fuzzy logic and artificial neural networks. The system works with variables from the process of the first circuit, second circuit and the containment and it is presented as a model for the integration of safety analyses in the process of decision making by the operator when tackling with accidental situations

  13. LTO License Application Project NPP Borssele

    International Nuclear Information System (INIS)

    Jong, A.E. de; Blom, F.J.; Leilich, J.

    2012-01-01

    Borssele NPP plans to extend its operating life with 20 years until 2034. Borssele has started the project LTO 'bewijsvoering' (LTO 'Justification') in order to meet the requirements of the Dutch regulator. The outline of the project is based on IAEA safety guide 57 'Safe Long Term Operation of Nuclear Power Plants'. This paper describes the contents and coherence of the different parts in the project and how these respond to the IAEA guidelines on LTO. The goal of the project LTO 'bewijsvoering' is to ensure that safety and safety relevant systems, structures and components continue to perform their intended functions during long term operation. The outcome of the project LTO 'bewijsvoering' will be used for a license change application and this will be submitted to the Dutch regulator KFD for approval of prolonged operation of Borssele NPP after 2013. (author)

  14. Britain's Nuclear Electric pursues a major programme to enhance safety

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    The UK state-owned nuclear utility, Nuclear Electric, is actively pursuing a major initiative inherited from its predecessor the Central Electricity Generating Board, known as the Station Enhancement Programme. The Enhancement Programme is a multi-million pound project and covers all Nuclear Electric's stations (including relevant aspects of the de-commissioning site at Berkeley). Its fundamental objective is to enhance the arrangements, practices and attitudes on the power stations to create an overall environment likely to promote ever safer and more reliable operation, and to take full account of the best international practices. The four year programme started in the spring of 1989, and is expected to be completed in 1993. Key elements of the programme include quality assurance based management systems; improvements to site and plant material conditions; a more formal approach to training; additional resources and controls to aid the management of maintenance defects; higher profile for the feedback of operational experience; management targets; a station evaluation programme; and formal feedback of lessons learned from the IAEA operational and safety review team visit to the Oldbury-on-Severn station in 1989. (author)

  15. Design of NPP of new generation being constructed at the Novovoronezh NPP site

    International Nuclear Information System (INIS)

    Afrov, A.; Berkovich, V.; Generalov, V.; Dragunov, Yu.; Krushelnitsky, V.

    1999-01-01

    The design of a new generation NPP is described, underscoring advances in physical attributes and passive safety systems based on experiences with earlier designs at operating NPPs. This paper elaborates on systems for handling and storing radioactive wastes, on refinements in containment measures and on experimental and analytic validation of critical design factors. (author)

  16. Training to Support Standardization and Improvement of Safety I and C Related Verification and Validation Activities

    Energy Technology Data Exchange (ETDEWEB)

    Ammon, G.; Schoenfelder, C.

    2014-07-01

    In recent years AREVA has conducted several measures to enhance the effectiveness of safety I and C related verification and validation activities within nuclear power plant (NPP) new build as well as modernization projects, thereby further strengthening its commitment to achieving the highest level of safety in nuclear facilities. (Author)

  17. Experimental analysis of specification language diversity impact on NPP software diversity

    International Nuclear Information System (INIS)

    Yoo, Chang Sik

    1999-02-01

    In order to increase computer system reliability, software fault tolerance methods have been adopted to some safety critical systems including NPP. Prevention of software common mode failure is very crucial problem in software fault tolerance, but the effective method for this problem is not found yet. In our research, to find out an effective method for prevention of software common mode failure, the impact of specification language diversity on NPP software diversity was examined experimentally. Three specification languages were used to compose three requirements specifications, and programmers made twelve product codes from the specifications. From the product codes analysis, using fault diversity criteria, we concluded that diverse specification language method would enhance program diversity through diversification of requirements specification imperfections

  18. Radiological impact on the surrounding area of Cernavoda NPP operation

    International Nuclear Information System (INIS)

    Busnita, M.; Penescu, M.; Neacsu, L.; Ion, M.; Moldoveanu, E.

    1996-01-01

    The objective of this paper is to present the protective measures provided to Cernavoda NPP for population and environment in case of normal operation and accidental situations. The paper describes the nuclear safety concept with the two aspects, the technical safety and the radiation protection, respectively, which represent the theoretical basis of the NPP design, construction and operation. The assessment of the potential radiological effects of Cernavoda NPP operation on the population and other environmental factors have been done by using the following input data: the natural and social characteristics of the site and the technical characteristics of the plant. The effects of Cernavoda NPP operation are also exposed in health risks which are compared with the daily risks of the population exposed of the natural phenomena or to the activities like transportation, domestics or industrial. (author). 1 tab., 6 refs

  19. Enhanced safety in the storage of fissile materials

    International Nuclear Information System (INIS)

    Williams, G.E.; Alvares, N.J.

    1978-01-01

    An inexpensive boron-loaded liner of epoxy resin for fissile-material storage containers was developed that can be easily fabricated of readily available, low-cost materials. Computer calculations indicate reactivity will be reduced substantially if this neutron-absorbing liner is added to containers in a typical storage array. These calculations compare favorably with neutron-attenuation experiments with thermal and fission neutron spectra, and tests at the Fire Test Facility indicate the epoxy resin will survive extreme environmental and accident conditions. The fire-resistant and insulating properties of the epoxy-resin liner further augment its ability to protect fissile materials. Boron-loaded epoxy resin is adaptable to many tasks but is particularly useful for providing enhanced criticality safety in the packaging and storage of fissile materials

  20. Distributed fiber optic sensing enhances pipeline safety and security

    Energy Technology Data Exchange (ETDEWEB)

    Frings, Jochen; Walk, Tobias [ILF Consulting Engineers, Munich (Germany)

    2011-09-15

    Pipelines are efficient, highly reliable and safe means of transportation. However, despite intensive right of way surveillance by foot, car and out of the air, pipeline leaks and illegal tappings are a reality - sometimes with catastrophic results. These events show a gap in real-time monitoring caused by the highly distributed nature of pipelines. Parts of this gap now can be closed with distributed fiber optic sensing technology. Using various physical effects this technology is apt to detect temperature, strain, vibrations and sound with very good localization over spans up to 50 km with a single sensor cable. Various field tested applications like leakage detection, third party activity monitoring and intrusion detection or ground movement detection as well as integrity monitoring proof that distributed fiber optic sensing can enhance pipeline safety and security. (orig.)

  1. The Alternative Design Features for Safety Enhancement in Shutdown Operation

    International Nuclear Information System (INIS)

    Oh, Hae Cheol; Kim, Myung Ki; Chung, Bag Soon; Seo, Mi Ro

    2009-01-01

    PSA can be used to confirm that the new plant design is complied with the applicable safety goals, and to select among the alternate design options. A shutdown PSA provides insight for outage planning schedule, outage management practices, and design modifications. Considering the results of both LPSD PSA studies and operating experiences for low power and shutdown, the improvements can be proposed to reduce the high risk contribution. The improvements/enhancements during shutdown operation may be divided into categories such as hardware, administrative management, and operational procedure. This paper presents on an example how the risk related to an accidental situation can be reduced, focusing the hardware design changes for the newly designed NPPs

  2. Contribution of the Slovak University of Technology Bratislava to the Education of NPP Operation Staff in Slovakia

    International Nuclear Information System (INIS)

    Hascik, J.; Slugen, V.; Hinca, R.; Miglierini, M.

    2006-01-01

    Paper is focused on the preparation of NPP VVER -440 staff in Slovak conditions. The realisation is managed via special technical courses, seminars, workshops, and trainings on selected experimental facilities at domestic as well as international level. Post-gradual re-qualification study: Safety aspects of NPP operation is discussed in detail. Six-year experience with NPP operating staff education can be shared and recommended also at international level. Based on these courses, special training for optimal preparation of NPP supervising physicists was started in 2002. In addition to all our activities, the international course: Safety aspects of NPP operation for subcontractors was prepared and realised in 2005.(author)

  3. Operating Experience at NPP Krsko

    International Nuclear Information System (INIS)

    Kavsek, D.; Bach, B.

    1998-01-01

    Systematic analysis of operational experience by assessment of internal and industry events and the feedback of lessons learned is one of the essential activities in the improvement of the operational safety and reliability of the nuclear power plant. At NPP Krsko we have developed a document called ''Operating Experience Assessment Program''. Its purpose is to establish administrative guidance for the processing of operating events including on-site and industry events. Assessment of internal events is based on the following methods: Event and Causal Factor Charting, Change Analysis, Barrier Analysis, MORT (Management Oversight and Risk Tree Analysis) and Human Performance Evaluation. The operating experience group has developed a sophisticated program entitled ''Operating experience tracking system'' (OETS) in response to the need for a more efficient way of processing internal and industry operating experience information. The Operating Experience Tracking System is used to initiate and track operational events including recommended actions follow up. Six screens of the system contain diverse essential information which allows tracking of operational events and enables different kinds of browsing. OETS is a part of the NPP Krsko nuclear network system and can be easily accessed by all plant personnel. (author)

  4. The NPP Isar comprehensive Aging Management Program

    International Nuclear Information System (INIS)

    Zander, Andre; Ertl, Stefan

    2012-01-01

    The majority of System, Structure and Components (SSC) in a nuclear power plants are designed to experience a service life, which is far above the intended design life. In most cases, only a small percentage of SSCs are subject to significant aging effects, which may affect the integrity or the function of the component. The process of aging management (AM) has the objective to monitor and control degradation effects which may compromise safety functions of the plant. And furthermore, to ensure, that testing and maintenance programs sufficiently provide preventive measures to control degradation effects. Safety-related aspects and the targeted high availability of the power plant as well as the requirements stipulated by German regulatory authorities prompted the operator of NPP ISAR to introduce an aging surveillance program. The NPP Isar as well as the German NPPs has to be following in the scope of aging management the KTA 1403 guideline. The NPP Isar surveillance program based on the KTA 1403 guideline covers the following aspects: - Scoping and screening of safety relevant Systems, Structures and Components (SSC); - Identification of possible degradation mechanisms for safety relevant SSC; - Ensure, that testing and maintenance programs sufficiently provide preventive measures to control degradation effects; - Transferability check of industry experience (internal and external events); - Annual preparation of an AM status report. (author)

  5. Technical innovations at NPP Dukovany - for safe and efficient operation

    International Nuclear Information System (INIS)

    Sabata, M.; Vasa, I.

    2000-01-01

    Inherent features of the NPP Dukovany design provide a significant confidence in its nuclear safety assurance; among these features should be emphasised the reactor core stability and its control and protection system capability to hold the reactor in safe state following scram or accident conditions. Nevertheless, NPP Dukovany was designed in the early seventies, and current requirements for nuclear safety assurance are more strict and/or specific as a result of the technical development and lessons learned from nuclear accidents. The paper compares the safety design base established at the time of NPP Dukovany project implementation and the current reference design base. The paper also presents procedures applied to implement technical and operational measures which are introduced to fulfil the current basic safety criteria. The scope of such measures applied at NPP Dukovany is compared with that of EU countries introduced for the same reason - to meet the updated safety related requirements. Examples of some innovations already implemented or under implementation give an idea how NPP Dukovany proceeds in reaching the goal of harmonising its safety with the requirements to be met before the Czech Republic becomes a member country of the European Union. (author)

  6. Engineering and Safety Partnership Enhances Safety of the Space Shuttle Program (SSP)

    Science.gov (United States)

    Duarte, Alberto

    2007-01-01

    difference for the safety of the Space Shuttle Vehicle, its crew, and personnel. Because of the MSERP's valuable contribution to the assessment of safety risk for the SSP, this paper also proposes an enhanced Panel concept that takes this successful partnership concept to a higher level of 'true partnership'. The proposed panel is aimed to be responsible for the review and assessment of all risk relative to Safety for new and future aerospace and related programs.

  7. Procedure for Application of Software Reliability Growth Models to NPP PSA

    International Nuclear Information System (INIS)

    Son, Han Seong; Kang, Hyun Gook; Chang, Seung Cheol

    2009-01-01

    As the use of software increases at nuclear power plants (NPPs), the necessity for including software reliability and/or safety into the NPP Probabilistic Safety Assessment (PSA) rises. This work proposes an application procedure of software reliability growth models (RGMs), which are most widely used to quantify software reliability, to NPP PSA. Through the proposed procedure, it can be determined if a software reliability growth model can be applied to the NPP PSA before its real application. The procedure proposed in this work is expected to be very helpful for incorporating software into NPP PSA

  8. Chernobyl NPP accident: a year later

    International Nuclear Information System (INIS)

    Asmolov, V.G.; Borovoj, A.A.; Demin, V.F.

    1988-01-01

    Consideration is being given to measures on liquidation of Chernobyl accident aftereffects, conducted since August, 1986. One of the most important measures lay in construction of the ''shelter'', which must provide long-term conservation of accidental unit. Works on decontamination of reactor area and contaminated populated regions were continued. Measures on providing safety of population and its health protection were performed. An attention was paid to long-term investigations on studying delayed aftereffects of the accident, monitoring of invironment, development and introduction of measures on improving NPP safety. Prospects of further development of nuclear power engeneering and possibilities of improving its safety are considered

  9. EMAS at Doel NPP

    International Nuclear Information System (INIS)

    Nolan, Peter; Thoelen, Els

    1998-01-01

    In October 1995, Doel NPP of Electrabel, Belgium opted to seek registration under the EC Eco-management and Audit Scheme (EMAS). A comprehensive environmental management system (EMS) has been introduced and implemented, encompassing all four PWRs and the supporting departments. A critical step was to seek certification from an accredited environmental auditing body against the International Standard ISO 14001. This provided the foundation for the publicly available environmental statement required by EMAS. The complications of achieving EMAS at a time when national and international standards were being re-formulated were successfully overcome and Doel NPP passed its EMAS audit in June 1997. (author)

  10. Application of Robotic System for Emergency Response in NPP

    International Nuclear Information System (INIS)

    Jeong, Kyung Min; Seo, Yong Chil; Shin, Ho Chul; Lee, Sung Uk; Cho, Jae Wan; Choi, Young Soo; Kim, Chang Hoi; Kim, Seung Ho

    2010-01-01

    Increasing energy demand and concerns over climate change make increasing use of nuclear power plant in worldwide. Even though the probability of accident is greatly reduced, safety is the highest priority issue in the nuclear energy industry. Applying highly reliable and conservative 'defense in depth' concepts with the design and construction of NPP, there are very little possibilities with which accidents are occur and radioactive materials are released to environments in NPP. But NPP have prepared with the emergency response procedures and conduct exercises for post-accident circumstance according to the procedures. The application of robots for emergency response task for post-accident in nuclear facilities is not a new concept. Robots have been sent to recover the damaged reactor at Chernobyl where human workers could receive a lifetime dose of radiation in minutes. Based on NRC's TMI-2 Cleanup Program, several robots were built in the 1980s to help gather information and remove debris from a reactor at the Three Mile Island nuclear power plant that partially melted down in 1979. The first robot was lowered into the basement through a hatch and human operators monitoring in a control room drove it through mud, water and debris, capturing the initial post-accident images of the reactor's basement. It was used for several years equipped with various tools allowing it to scour surfaces, scoop samples and vacuum sludge. A second version carried a core sampler to determine the intensity and depth of the radiation that had permeated into the walls. To perform cleanup tasks, they built Workhorse that featured system redundancy and had a boom extendable to reach high places, but it was never used because it had too many complexities and to clean and fix. While remote robotics technology has proven to remove the human from the radioactive environment, it is also difficult to make it useful because it may requires skill about remote control and obtaining remote

  11. Application of Robotic System for Emergency Response in NPP

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyung Min; Seo, Yong Chil; Shin, Ho Chul; Lee, Sung Uk; Cho, Jae Wan; Choi, Young Soo; Kim, Chang Hoi; Kim, Seung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    Increasing energy demand and concerns over climate change make increasing use of nuclear power plant in worldwide. Even though the probability of accident is greatly reduced, safety is the highest priority issue in the nuclear energy industry. Applying highly reliable and conservative 'defense in depth' concepts with the design and construction of NPP, there are very little possibilities with which accidents are occur and radioactive materials are released to environments in NPP. But NPP have prepared with the emergency response procedures and conduct exercises for post-accident circumstance according to the procedures. The application of robots for emergency response task for post-accident in nuclear facilities is not a new concept. Robots have been sent to recover the damaged reactor at Chernobyl where human workers could receive a lifetime dose of radiation in minutes. Based on NRC's TMI-2 Cleanup Program, several robots were built in the 1980s to help gather information and remove debris from a reactor at the Three Mile Island nuclear power plant that partially melted down in 1979. The first robot was lowered into the basement through a hatch and human operators monitoring in a control room drove it through mud, water and debris, capturing the initial post-accident images of the reactor's basement. It was used for several years equipped with various tools allowing it to scour surfaces, scoop samples and vacuum sludge. A second version carried a core sampler to determine the intensity and depth of the radiation that had permeated into the walls. To perform cleanup tasks, they built Workhorse that featured system redundancy and had a boom extendable to reach high places, but it was never used because it had too many complexities and to clean and fix. While remote robotics technology has proven to remove the human from the radioactive environment, it is also difficult to make it useful because it may requires skill about remote control and

  12. Computer-assisted navigational surgery enhances safety in dental implantology.

    Science.gov (United States)

    Ng, F C; Ho, K H; Wexler, A

    2005-06-01

    Dental implants are increasingly used to restore missing dentition. These titanium implants are surgically installed in the edentulous alveolar ridge and allowed to osteointegrate with the bone during the healing phase. After osseo-integration, the implant is loaded with a prosthesis to replace the missing tooth. Conventional implant treatment planning uses study models, wax-ups and panoramic x-rays to prefabricate surgical stent to guide the preparation of the implant site. The drilling into the alveolar ridge is invariably a "blind" procedure as the part of the drill in bone is not visible. Stereotactic systems were first introduced into neurosurgery in 1986. Since then, computer-assisted navigational technology has brought major advances to neuro-, midface and orthopaedic surgeries, and more recently, to implant placement. This paper illustrates the use of real-time computer-guided navigational technology in enhancing safety in implant surgical procedures. Real-time computer-guided navigational technology enhances accuracy and precision of the surgical procedure, minimises complications and facilitates surgery in challenging anatomical locations.

  13. The Hungarian model project: Strengthening training for operational safety at Paks nuclear power plant

    International Nuclear Information System (INIS)

    Mautner Markhof, F.

    1998-01-01

    The Hungarian Model project (HMP) reflects the commitment to constant increase of safety and reliability of the NPP Paks, the Government of Hungary and the IAEA. It includes some of the most important nuclear power objectives of Paks NPP, namely the strengthening of NPP personnel training and competence through the application of international best practice, the systematic approach to training (SAT), for training operation and maintenance personnel; setting up a state of-the-art maintenance training center (MTC) at Paks and enhancing safety culture at Paks NPP. The IAEA supported implementation of the HMP through fellowships and scientific visits, expert missions, provision of hardware and software for SAT application, and supply od major new uncontaminated items of actual WWER equipment for the MTC

  14. Report of the review of the safety improvement programme for South Ukraine NPP units 1 and 2 and to identify the safety issues of ''small series'' WWER-1000 NPPs. South Ukraine Yuzhnoukrainsk, Nikolaev Region Ukraine, 8 to 19 July 1996. Draft

    International Nuclear Information System (INIS)

    Bastin, S.; Hoehn, J.; Lin, C.; Taylor, R.; Benitez, F.; Dale, H.; Mueller, B.; Rieg, C.Y.

    1996-10-01

    According to the Ukrainian request the purpose of the IAEA experts' mission was to review the safety improvement programme for South Ukraine NPP Units 1 and 2 in order to advise on the completeness and adequacy of safety improvements implemented and/or proposed. Another purpose of the mission was to identify major design and operational deficiencies as a basis to compile a consolidated list of generic safety issues for the units of the 'small series'' of WWER-1000 reactors (''Issue Book for ''small series'' WWER-1000 NPPs). Conclusions and recommendations from the IAEA mission are based on the combined expertise of the international group of experts who composed the team. They are intended to assist national authorities and plant operators who have the sole responsibilities for the regulation and safe operation. tabs

  15. Aspects of accident management in Cernavoda NPP

    International Nuclear Information System (INIS)

    Dascalu, N.

    1999-01-01

    As a general conclusion, the accident management system as implemented at Cerna voda NPP is expected to be appropriate for handling a severe accident, should it occur, in such a way that the environmental radiological consequences would be insignificant and radiation exposure of the personnel be within recommendations. It is recognized, however, that continued development and verification of the system as well as effective personnel training programs are essential to maintain the safety level achieved. (author)

  16. Considerations related to Cernavoda NPP lifetime management

    International Nuclear Information System (INIS)

    Cojan, Mihail

    2007-01-01

    Cernavoda NPP, the first CANDU in Europe, is one of the original CANDU 6 plants and the first CANDU 6 producing over 706 MWe. While the first series of CANDU 6 plants (which entered service in the early 1980 s) have now reached the 2/3 of their 30 years design life, the Cernavoda NPP was put into service on the 2nd December 1996. After 10 years of operation the Plant Life Management (PLiM) Program for Cernavoda should be an increasingly important program to Utility ('CNE - Prod') in order to protect the investment and the continued success of plant operation. The goal of the paper is to present some considerations related to Cernavoda NPP lifetime management. The Plant Life Management Program, known as PLiM Program is concerned with the analysis of technical limits of the safe operation - from the point of view of nuclear safety - in NPP units, aiming at attaining the planned 30 years life duration and its extension to 40 or even 50 years of safe and economical operation. For the CANDU reactors the so-called PLiM and PLEX Programs are just applied. These are applied research programs that approach with priority the current practices for assessing the capability of safe operation within the limits of nuclear safety (fitness-for-service assessment). These programs also approach inspection, monitoring are prevention of degrading due to the ageing of critical systems, structures and components (CSSCs). As each nuclear plant is somewhat different in its components and systems, materials composition, procurement, construction, and operational history, directed research and development programs into materials behavior, monitoring techniques, and methods to mitigate ageing are required to support the lifetime management. Over the past 6 years, INR Pitesti (Institute for Nuclear Research - Romania) has been working on R and D Programs to support a comprehensive and integrated Cernavoda NPP Life Management Program (PLiM) that will see the Cernavoda NPP successfully and

  17. Using Total Lightning Observations to Enhance Lightning Safety

    Science.gov (United States)

    Stano, Geoffrey T.

    2012-01-01

    Lightning is often the underrated threat faced by the public when it comes to dangerous weather phenomena. Typically, larger scale events such as floods, hurricanes, and tornadoes receive the vast majority of attention by both the general population and the media. This comes from the fact that these phenomena are large, longer lasting, can impact a large swath of society at one time, and are dangerous events. The threat of lightning is far more isolated on a case by case basis, although millions of cloud-to-ground lightning strikes hit this United States each year. While attention is given to larger meteorological events, lightning is the second leading cause of weather related deaths in the United States. This information raises the question of what steps can be taken to improve lightning safety. Already, the meteorological community s understanding of lightning has increased over the last 20 years. Lightning safety is now better addressed with the National Weather Service s access to the National Lightning Detection Network data and enhanced wording in their severe weather warnings. Also, local groups and organizations are working to improve public awareness of lightning safety with easy phrases to remember, such as "When Thunder Roars, Go Indoors." The impacts can be seen in the greater array of contingency plans, from airports to sports stadiums, addressing the threat of lightning. Improvements can still be made and newer technologies may offer new tools as we look towards the future. One of these tools is a network of sensors called a lightning mapping array (LMA). Several of these networks exist across the United States. NASA s Short-term Prediction Research and Transition Center (SPoRT), part of the Marshall Spaceflight Center, has access to three of these networks from Huntsville, Alabama, the Kennedy Space Center, and Washington D.C. The SPoRT program s mission is to help transition unique products and observations into the operational forecast environment

  18. Temelin NPP commissioning experiences

    International Nuclear Information System (INIS)

    Hanus, V.

    2002-01-01

    The Building Permit for the Temelin NPP with four VVER units was issued in 1986, which is a long time ago. Since then, however, was taken a route that is very different from what anybody imagined. Described are the legislative and design changes and given is a current condition of the power plant

  19. NPP Krsko decommissioning concept

    International Nuclear Information System (INIS)

    Novsak, M.; Fink, K.; Spiler, J.

    1996-01-01

    At the end of the operational lifetime of a nuclear power plant (NPP) it is necessary to take measures for the decommissioning as stated in different international regulations and also in the national Slovenian law. Based on these requirements Slovenian authorities requested the development of a site specific decommissioning plan for the NPP KRSKO. In September 1995, the Nuklearna Elektrarna Krsko (NEK) developed a site specific scope and content for decommissioning plan including the assumptions for determination of the decommissioning costs. The NEK Decommissioning Plan contains sufficient information to fulfill decommissioning requirements identified by NRC, IAEA and OECD - NEA regulations. In this paper the activities and the results of development of NEK Decommissioning Plan consisting of the development of three decommissioning strategies for the NPP Krsko and selection of the most suitable strategy based on site specific, social, technical, radiological and economical aspects, cost estimates for the strategies including the costs for construction of final disposal facilities for fuel/high level waste (fuel/HLW) and low/intermediate level waste (LLW/ILW) and scheduling all activities necessary for the decommissioning of the NPP KRSKO are presented. (author)

  20. NPP Krsko decommissioning concept

    International Nuclear Information System (INIS)

    Novsak, M.; Fink, K.; Spiler, J.

    1996-01-01

    At the end of the operational lifetime of a nuclear power plant (NPP) it is necessary to take measures for the decommissioning as stated in different international regulations and also in the national Slovenian law. Based on these requirements Slovenian authorities requested the development of a site specific decommissioning plan for the NPP Krsko. In September 1995, the Nuklearna Elektrarna Krsko (NEK) developed a site specific scope and content for a decommissioning plan including the assumptions for determination of the decommissioning costs. The NEK Decommissioning Plan contains sufficient information to fulfill the decommissioning requirements identified by NRC, IAEA and OECD - NEA regulations. In this paper the activities and results of development of NEK Decommissioning Plan consisting of the development of three decommissioning strategies for the NPP Krsko and selection of the most suitable strategy based on site specific, social, technical, radiological and economic aspects, cost estimates for the strategies including the costs for construction of final disposal facilities for fuel/high level waste (fuel/HLW) and low/intermediate level waste (LLW/ILW) and scheduling of all activities necessary for the decommissioning of the NPP Krsko are presented. (author)

  1. Can the Ignalina NPP be safe?

    International Nuclear Information System (INIS)

    Kutas, S.

    1999-01-01

    Many countries started using nuclear power before Lithuania, the IAEA and NEA were established and have been functioning efficiently, numerous conventions have been signed that control nuclear activities, nuclear safety regulations and standards have been set. Lithuania which is striving to integrate into the West European structures and to become a member of the 'club' of the world's nuclear states, has become a full and equal member of the IAEA and recognises its recommendations. Furthermore, it has signed or joined international conventions and is doing its best to introduce the practice of controlling nuclear safety applied in western countries. Nuclear safety cannot be taken as something finite that does not change. The requirements of nuclear and radiation safety are continuously redefined taking into consideration scientific discoveries and inventions as well as technical progress. Safety assurance can be the only criterion that should determine whether the nuclear power plant can be operated in Lithuania. The safety analysis report confirmed that the Ignalina NPP can be safely operated until the channel-graphite gap closes. The experiments conducted to date have convincingly shown that by the year 2000 the gap will not close in any of the fuel channels of unit 1. It can be hoped that the submitted studies, analyses and assessments will confirm that the Ignalina NPP meets Lithuania's nuclear safety requirements that have been drawn up in compliance with the recommendations of IAEA

  2. Two managerial grids in NPP

    International Nuclear Information System (INIS)

    Zhao Hui

    2012-01-01

    Today, the nuclear power corporation (NPC) enjoys the profit of LCEP (the low carbon economic policy). at the same time, they also enduring more and more pressure. For example, the partner competition or the NPP potential occupational risk . The efficient counterplot of risk is the self-ability cultivation. It is essential to research the NPP managerial flow. The nuclear power plant (NPP) unit is a carrier of the NPC enterprise management system, and has taken on a new look 'pull one portion then the whole moving'. The NPP has three systematical characters, the security responsibility center, the man-machine system and the input-output system. The manufacturing system and the enterprise management system are the great constituents of the NPP managerial flows. Means of systems analysis, we can find out the truth of the NPP running interface. In CHINA, there are many operating experiences near 20 years. It indicates that the NPP manufacturing system and the enterprise system are the roots of the nuclear power corporation, the core of the all NPP systems must be based on it. So the ability cultivation is the work core to NPP. It is reliably to ensure the NPP to be up against problems, for instance, the security duty, the costing control and the man-machine system running harmoniously. This paper introduces the NPP managerial flow and the present state of QNPC, also come up with a proposal to refer for the NPC development actions of collective measure, specialization, standardization, fine. (author)

  3. Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Baji, C.; Vamos, G.; Toth, J.

    2001-01-01

    The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

  4. Some aspects of RPV integrity of Ukrainian NPP's

    International Nuclear Information System (INIS)

    Zaritsky, N.; Kovyrshin, V.; Zhukov, P.

    1998-01-01

    The operating organisations in Ukraine implement the main IAEA recommendations concerned with NPP operational safety. Sufficient substantiation of the measures for improvement of WWER-1000 NPPs safety and reliability is provided. General information (chemical compositions and mechanical properties of reactor pressure vessel materials) is already collected on all Ukrainian WWER-1000 NPPs and transferred to International database on Pressure Vessel materials

  5. The post-accidents exploitation of Chernobyl NPP

    International Nuclear Information System (INIS)

    Nosovskij, A.V.

    1996-01-01

    Problems facing the Chernobyl NPP personnel since its commissioning and during its post-accident operation are considered. The accomplished measures on improving the power units safety, normalization of the radiation situation as well as the impact of psychological factor at the safety culture level are discussed

  6. Situation analysis of physical independence of the equipment and safety circuits of Almaraz NPP regarding R.G. 1.75 rev.3 (2005)

    International Nuclear Information System (INIS)

    Seijas Portela, S.

    2010-01-01

    Situation analysis of physical independence of the electrical equipment and circuits CN safety Almaraz about R.G. 1.75 rev. 3. (2005) The aim of this paper is to present the work done in the analysis of the physical separation of redundant safety electrical equipment (emergency diesel generators, medium voltage, electrical cabinets, etc.) and physical separation of circuits and electrical conduits.

  7. Optimization of design solutions on safety and economy for power unit of NPP with WWER reactor of new generation. Annex 8

    International Nuclear Information System (INIS)

    Krushelnitsky, V.N.; Berkovich, V.M.; Shvyrayev, Yu.; Podshebaykin, A.K.; Fil, N.S.

    2002-01-01

    Development of new generation WWER reactors is being carried out in Russia. These new projects with WWER reactors aim to achieve increased levels of safety and reduced costs. This paper describes these designs and discusses the main factors leading to the safety level increase and the improved economics. (author)

  8. Revision of Krsko NPP Quality Assurance Plan

    International Nuclear Information System (INIS)

    Biscan, R.; Fifnja, I.; Kavsek, D.

    2012-01-01

    International standards from nuclear power plant operation area are being frequently upgraded and revised in accordance with the continuous improvement philosophy. This philosophy applies also to the area of Quality Assurance, which has also undergone significant improvement since the early 1950s. Besides just nuclear industry, there are also other international quality standards that are being continuously developed and revised, bringing needs for upgrades also in the nuclear application. Since the beginning of Krsko NPP construction, the overall Quality Assurance program and its applicable procedures were in place to assure that all planned and systematic actions necessary to provide adequate confidence that an item or service will satisfy given requirements to quality, are in place. The overall requirements for quality as one of the major objectives for Krsko NPP operation are also set forth in the Updated Safety Analyses Report, the document that serves as a base for operating license. During more than 30 years of Krsko NPP operation, the quality requirements and related documents were revised and upgraded in several attempts. The latest revision 6 of QD-1, Quality Assurance Plan was issued during the year 2011. The bases for the revision were: Changes of the Slovenian regulatory requirements (ZVISJV, JV5, JV9?), Changes of Krsko NPP licensing documents (USAR section 13?), SNSA inspection requirements, Changes of international standards (IAEA, ISO?), Conclusions of first PSR, Implementation of ISO standards in Krsko NPP (ISO14001, ISO17025), Changes of plant procedures, etc. One of the most obvious changes was the enlargement of the QA Plan scope to cover interdisciplinary areas defined in the plant management program MD-1, such as Safety culture, Self-assessment, Human performance, Industrial Safety etc. The attachment of the QA Plan defining relationships between certain standards was also updated to provide matrix for better correlation of requirements of

  9. Food safety security: a new concept for enhancing food safety measures.

    Science.gov (United States)

    Iyengar, Venkatesh; Elmadfa, Ibrahim

    2012-06-01

    The food safety security (FSS) concept is perceived as an early warning system for minimizing food safety (FS) breaches, and it functions in conjunction with existing FS measures. Essentially, the function of FS and FSS measures can be visualized in two parts: (i) the FS preventive measures as actions taken at the stem level, and (ii) the FSS interventions as actions taken at the root level, to enhance the impact of the implemented safety steps. In practice, along with FS, FSS also draws its support from (i) legislative directives and regulatory measures for enforcing verifiable, timely, and effective compliance; (ii) measurement systems in place for sustained quality assurance; and (iii) shared responsibility to ensure cohesion among all the stakeholders namely, policy makers, regulators, food producers, processors and distributors, and consumers. However, the functional framework of FSS differs from that of FS by way of: (i) retooling the vulnerable segments of the preventive features of existing FS measures; (ii) fine-tuning response systems to efficiently preempt the FS breaches; (iii) building a long-term nutrient and toxicant surveillance network based on validated measurement systems functioning in real time; (iv) focusing on crisp, clear, and correct communication that resonates among all the stakeholders; and (v) developing inter-disciplinary human resources to meet ever-increasing FS challenges. Important determinants of FSS include: (i) strengthening international dialogue for refining regulatory reforms and addressing emerging risks; (ii) developing innovative and strategic action points for intervention {in addition to Hazard Analysis and Critical Control Points (HACCP) procedures]; and (iii) introducing additional science-based tools such as metrology-based measurement systems.

  10. Public safety investigations-A new evolutionary step in safety enhancement?

    International Nuclear Information System (INIS)

    Stoop, John; Roed-Larsen, Sverre

    2009-01-01

    A historical overview highlights the evolutionary nature of developments in accident investigations in the transport industry. Based on a series of major events outside transportation, the concept of accident investigations has broadened to other domains and to a widening of the scope of the investigation. Consequently, existing investigation boards are forced to adapt their mandates, missions and methods. With the introduction of social risk perception and application of the concept of safety investigation in the public sector, a change of focus towards the aftermath and non-technical issues of a more generic nature emerges. This expansion has also gained the interest of social sciences and public governance, generating new underlying models and theories on risk and responsibility. The evolutionary development of safety investigations is demonstrated by the various organisational forms which shaped accident investigations in different countries. Underneath these organisational differences, a need for a common methodology and a reflection on fundamental notions is discussed. In particular differences among human operator models, the allocation of responsibilities in design concepts and methodological issue are elaborated. The needs and opportunities for a transition from accident prevention towards systems change are indicated. At present, the situation is ambiguous. An encompassing inventory can only provide a general oversight over emerging trends and lacks analytic rigor on specific topics. The societal dimensions, institutional changes at the level of governance and control and the powers that advocate or challenge investigations are not yet fully described. Therefore, in the conclusions a small number of critical challenges and threats are identified that should be open to scrutiny in order to facilitate a new, evolutionary step in safety enhancement.

  11. Operating procedure automation to enhance safety of nuclear power plants

    International Nuclear Information System (INIS)

    Husseiny, A.A.; Sabri, Z.A.; Adams, S.K.; Rodriguez, R.J.; Packer, D.; Holmes, J.W.

    1989-01-01

    Use of logic statements and computer assist are explored as means for automation and improvement on design of operating procedures including those employed in abnormal and emergency situations. Operating procedures for downpower and loss of forced circulation are used for demonstration. Human-factors analysis is performed on generic emergency operating procedures for three strategies of control; manual, semi-automatic and automatic, using standard emergency operating procedures. Such preliminary analysis shows that automation of procedures is feasible provided that fault-tolerant software and hardware become available for design of the controllers. Recommendations are provided for tests to substantiate the promise of enhancement of plant safety. Adequate design of operating procedures through automation may alleviate several major operational problems of nuclear power plants. Also, automation of procedures is necessary for partial or overall automatic control of plants. Fully automatic operations are needed for space applications while supervised automation of land-based and offshore plants may become the thrust of new generation of nulcear power plants. (orig.)

  12. Acoustic Waves: A Route to Enhance Sodium Fast Reactor Safety

    International Nuclear Information System (INIS)

    Jeannot, Jean-Philippe; Baque, François; Cavaro, Matthieu; Gastaldi, Olivier; Lhuilier, Christian; Massacret, Nicolas; Moriot, Jérémy; Paumel, Kévin; Vandergaegen, Matthias; Rodriguez, Gilles

    2013-01-01

    Improvement to prevent core meltdown and to provide a more robust safety demonstration → Safety objectives: - A level of safety at least equivalent to EPR’s level, - Consolidation of the defence-in-depth principle, - A more robust safety demonstration than those of the Phenix and Superphenix reactor. Acoustic techniques: - Low attenuation by the sodium medium - High velocity of US wave (2289 m.s-1 at 550°C) →

  13. Assessment on regulatory framework for safety of radioactive waste management as preparation for future NPP in Indonesia; a comparison study with South Korea

    International Nuclear Information System (INIS)

    Setianingsih, Lilis Susanti; Jeong, Seung Young

    2012-01-01

    Indonesia is an archipelago country consisting of more than 18,000 islands with five main big islands; Java; the smallest among the five, with the most population which is more than 58% of the total 237 million people as national population living in it; Sumatera, Borneo/Kalimantan, Sulawesi and Irian which share half of its island to Papua New Guinea. Challenges in monitoring and controlling nuclear related facilities in such widely spread geographical location is definitely high. Requirements of strong regulation and consistent law enforcement along with qualified human resources to perform their duties need to be satisfied in order to be able to cover the whole region with excellent strategic planning of actions. Apart from the three research reactors, radioactive and/or nuclear facilities exist in industrial sites in Indonesia for various purposes. There are several types of nuclear energy utilization for industrial purposes. Irradiator is one type utilizing ionizing radiation generation or sealed encapsulated radioactive source commonly used for enhancing rubber flexibility without decreasing its quality level. Other application of nuclear energy utilization in industry is gauging implemented in beverage and cigarettes industries for measurement process by using radioactive source like Cs 137, Co 60 and Am 241Be to control maximum level of liquid in glass bottles and to check even distribution of tobacco in cigarettes. Nuclear energy utilization in Non Destructive Testing (NDT) application can be found in facilities like radiography for open facilities (usually for remote location with mobile equipment of NDT) and for closed facilities usually performed for piping inspection particularly in piping connection for structure and material inspection. In oil field service industry, we have oil well logging utilizing nuclear energy for formation evaluation process by having porosity and density measurements. Nuclear well logging is intended to get well bore

  14. Assessment on regulatory framework for safety of radioactive waste management as preparation for future NPP in Indonesia; a comparison study with South Korea

    Energy Technology Data Exchange (ETDEWEB)

    Setianingsih, Lilis Susanti [KAIST, Daejeon (Korea, Republic of); Jeong, Seung Young [KINS, Daejeon (Korea, Republic of)

    2012-10-15

    Indonesia is an archipelago country consisting of more than 18,000 islands with five main big islands; Java; the smallest among the five, with the most population which is more than 58% of the total 237 million people as national population living in it; Sumatera, Borneo/Kalimantan, Sulawesi and Irian which share half of its island to Papua New Guinea. Challenges in monitoring and controlling nuclear related facilities in such widely spread geographical location is definitely high. Requirements of strong regulation and consistent law enforcement along with qualified human resources to perform their duties need to be satisfied in order to be able to cover the whole region with excellent strategic planning of actions. Apart from the three research reactors, radioactive and/or nuclear facilities exist in industrial sites in Indonesia for various purposes. There are several types of nuclear energy utilization for industrial purposes. Irradiator is one type utilizing ionizing radiation generation or sealed encapsulated radioactive source commonly used for enhancing rubber flexibility without decreasing its quality level. Other application of nuclear energy utilization in industry is gauging implemented in beverage and cigarettes industries for measurement process by using radioactive source like Cs 137, Co 60 and Am 241Be to control maximum level of liquid in glass bottles and to check even distribution of tobacco in cigarettes. Nuclear energy utilization in Non Destructive Testing (NDT) application can be found in facilities like radiography for open facilities (usually for remote location with mobile equipment of NDT) and for closed facilities usually performed for piping inspection particularly in piping connection for structure and material inspection. In oil field service industry, we have oil well logging utilizing nuclear energy for formation evaluation process by having porosity and density measurements. Nuclear well logging is intended to get well bore

  15. Preparation of Long Term Operation in Dukovany NPP, Czech Republic

    International Nuclear Information System (INIS)

    Krivanek, R.; Sabata, M.

    2012-01-01

    Dukovany NPP in the south-east of the Czech Republic operates four VVER 440/213 type units. The first unit was commissioned in 1985 and the last one in 1987. The operational results of the whole NPP have been excellent and NPP permanently belongs between the first quartile of the best operated NPPs in the world in accordance with WANO factors. Large safety improvement programme have been implemented in last 15 years. The original design lifetime of main components is 30 years which means till 2015 and it is understandable that NPP is preparing for long-term operation (LTO). The paper is describing activities carried out and planned for safe and successful LTO. (author)

  16. Review of accident analyses performed at Mochovce NPP

    International Nuclear Information System (INIS)

    Siko, D.

    2000-01-01

    In this paper the review of accident analysis performed in NPP Mochovce V-1 is presented. The scope of these safety measures was defined and development in the T SSM for NPP Mochovce Nuclear Safety Improvements Report' issued in July 1995. The main objectives of these safety measures were the followings: (a) to establish the criteria for selection and classification of accidental events, as well as defining the list of initiating events to be analysed. Accident classification to the individual groups must be performed in accordance with RG 1.70 and IAEA recommendations 'Guidelines for Accidental Analysis of WWER NPP' (IAEA-EBR-WWER-01) to select boundary cases to be calculated from the scope of initiating events; (b ) to elaborate the accident analysis methodology that also includes acceptance criteria for their result evaluation, initial and boundary conditions, assumption related with the application of the single failure criteria, requirements on the analysis quality, used computer codes, as well as NPP models and input data for the accident analysis; (c) to perform the accident analysis for the Pre-operational Safety Report (POSAR); (d) to provide a synthetic report addressing the validity range of codes models and correlations, the assessment against relevant tests results, the evidence of the user qualification, the modernisation and nodding scheme for the plant and the justification of used computer codes. Analyses results showed that all acceptance criteria were met with satisfactory margin and design of the NPP Mochovce is accurate. (author)

  17. Chernobyl NPP decommissioning efforts - Past, Present and Future. Decommissioning Efforts on Chernobyl NPP site - Past, Present

    International Nuclear Information System (INIS)

    Kuchinskiy, V.

    2017-01-01

    Two unique large-scale projects are underway at the moment within the Chernobyl - Exclusion zone - Shelter object transformation into ecologically safe system and the decommissioning of 3 Chernobyl NPP Units. As a result of beyond design accident in 1986 the entire territory of the industrial site and facilities located on it was heavily contaminated. Priority measures were carried out at the damaged Unit under very difficult conditions to reduce the accident consequences and works to ensure nuclear and radiation safety are continuous, and the Unit four in 1986 was transformed into the Shelter object. Currently, works at the Shelter object are in progress. Under assistance of the International Community new protective construction was built above the existing Shelter object - New Safe Confinement, which will ensure the SO Safety for the long term - within up to 100 years. The second major project is the simultaneous decommissioning of Chernobyl NPP Units 1, 2 and 3. Currently existing Chernobyl NPP decommissioning Strategy has been continuously improved starting from the Concept of 1992. Over the years the following was analyzed and taken into account: the results of numerous research and development works, international experience in decommissioning, IAEA recommendations, comments and suggestions from the governmental and regulatory bodies in the fields of nuclear energy use and radioactive waste management. In 2008 the final decommissioning strategy option for Chernobyl NPP was approved, that was deferred gradual dismantling (SAFSTOR). In accordance with this strategy, decommissioning will be carried out in 3 stages (Final Shutdown and Preservation, Safe Enclosure, Dismantling). The SAFSTOR strategy stipulates: -) the preservation of the reactor, the primary circuit and the reactor compartment equipment; -) the dismantling of the equipment external in relation to the reactor; -) the safe enclosure (under the supervision); -) the gradual dismantling of the primary

  18. 76 FR 40648 - Safety Enhancements Part 139, Certification of Airports; Reopening of Comment Period

    Science.gov (United States)

    2011-07-11

    ... that was published on February 1, 2011. In that document, the FAA proposed several safety enhancements...-0247; Notice No. 11-01] RIN 2120-AJ70 Safety Enhancements Part 139, Certification of Airports... comment period for the NPRM published on February 1, 2011 (76 FR 5510) and reopened (76 FR 20570) April 13...

  19. Operating experience and systems analysis at Trillo NPP: A program intended for systematic review of plant safety systems to assess design basis requirements compliance

    International Nuclear Information System (INIS)

    Vega, R. de la

    1996-01-01

    The program was defined to apply to all plant safety systems and/or systems included in plant Technical Specifications. The goal of the program was to ensure, by systematic design, construction, and commissioning review, the adequacy of safety systems, structures and components to fulfill their safety functions. Also, as a result of the program, it was established that a complete, unambiguous, systematic, design basis definition shall take place. And finally, a complete documental review of the plant design shall result from the program execution

  20. Core design methodology and software for Temelin NPP

    International Nuclear Information System (INIS)

    Havluj, F; Hejzlar, J.; Klouzal, J.; Stary, V.; Vocka, R.

    2011-01-01

    In the frame of the process of fuel vendor change at Temelin NPP in the Czech Republic, where, starting since 2010, TVEL TVSA-T fuel is loaded instead of Westinghouse VVANTAGE-6 fuel, new methodologies for core design and core reload safety evaluation have been developed. These documents are based on the methodologies delivered by TVEL within the fuel contract, and they were further adapted according to Temelin NPP operational needs and according to the current practice at NPP. Along with the methodology development the 3D core analysis code ANDREA, licensed for core reload safety evaluation in 2010, have been upgraded in order to optimize the safety evaluation process. New sequences of calculations were implemented in order to simplify the evaluation of different limiting parameters and output visualization tools were developed to make the verification process user friendly. Interfaces to the fuel performance code TRANSURANUS and sub-channel analysis code SUBCAL were developed as well. (authors)

  1. Computer Security of NPP Instrumentation and Control Systems: Cyber Threats

    International Nuclear Information System (INIS)

    Klevtsov, A.L.; Trubchaninov, S.A.

    2015-01-01

    The paper is devoted to cyber threats, as one of the aspects in computer security of instrumentation and control systems for nuclear power plants (NPP). The basic concepts, terms and definitions are shortly addressed. The paper presents a detailed analysis of potential cyber threats during the design and operation of NPP instrumentation and control systems. Eleven major types of threats are considered, including: the malicious software and hardware Trojans (in particular, in commercial-off-the-shelf software and hardware), computer attacks through data networks and intrusion of malicious software from an external storage media and portable devices. Particular attention is paid to the potential use of lower safety class software as a way of harmful effects (including the intrusion of malicious fragments of code) on higher safety class software. The examples of actual incidents at various nuclear facilities caused by intentional cyber attacks or unintentional computer errors during the operation of software of systems important to NPP safety.

  2. Modernization of Unit 2 at Oskarshamn NPP- Main Objectives, Experience from Design, Separation of Operational and Nuclear Safety Equipment - Lessons Learned

    International Nuclear Information System (INIS)

    Kanaan, Salah K.

    2015-01-01

    This paper aims to give a picture of Oskarshamn Nuclear Power Plant (OKG) experience from design for one of the biggest modernization project in the world and focuses on what was learned that is specific to robustness of electrical power systems, especially through Fukushima Station Blackout (SBO). The planning for unit 2 at OKG was initiated in 1967 and the plant was completed on time and was synchronized to the grid October 2, 1974 and is of type BWR. Unit 2 was originally on 580 MW. In 1982 a thermal power up-rate was performed, from 1700 MWh to 1800 MWh (106% reactor output). A decision was made to perform a modernization and a new power up-rate to 850 MW and there were several reasons for this decision; New safety regulations from Swedish Radiation Safety Authority (SSM), Ageing of important components and the initial focus was on safety and availability - Project Plant Life Extension (Plex) was established and became the largest nuclear power modernization in the world. The modernization will lead to: - New safety concept with 4 divisions instead for existing 2 with 2 new buildings South Electrical Building (SEB) and North Electrical Building (NEB); - Completely new software - based equipments for monitoring, control and I and C; - New Low Pressure Turbine, new generator and main transformer; - New MCR and simulator; - Compliance with modern reactor safety requirements; - Redundancy, Separation, Diversification, Earthquake; - Reinforcement of existing safety functions; - New Electricity - I and C (electric power incl. reinforced emergency power and control systems); - New buildings for Electricity - I and C; - Reinforcement of existing process systems as well as installation of new ones. Based on studies and good experiences on how to separate the operational and the safety equipment, the project led to a completely new safety concept. The safety concept is based on fully understanding the safety system that shall encompass all of the elements required to

  3. Status of the NPP and J1 NOAA Unique Combined Atmospheric Processing System (NUCAPS): recent algorithm enhancements geared toward validation and near real time users applications.

    Science.gov (United States)

    Gambacorta, A.; Nalli, N. R.; Tan, C.; Iturbide-Sanchez, F.; Wilson, M.; Zhang, K.; Xiong, X.; Barnet, C. D.; Sun, B.; Zhou, L.; Wheeler, A.; Reale, A.; Goldberg, M.

    2017-12-01

    The NOAA Unique Combined Atmospheric Processing System (NUCAPS) is the NOAA operational algorithm to retrieve thermodynamic and composition variables from hyper spectral thermal sounders such as CrIS, IASI and AIRS. The combined use of microwave sounders, such as ATMS, AMSU and MHS, enables full atmospheric sounding of the atmospheric column under all-sky conditions. NUCAPS retrieval products are accessible in near real time (about 1.5 hour delay) through the NOAA Comprehensive Large Array-data Stewardship System (CLASS). Since February 2015, NUCAPS retrievals have been also accessible via Direct Broadcast, with unprecedented low latency of less than 0.5 hours. NUCAPS builds on a long-term, multi-agency investment on algorithm research and development. The uniqueness of this algorithm consists in a number of features that are key in providing highly accurate and stable atmospheric retrievals, suitable for real time weather and air quality applications. Firstly, maximizing the use of the information content present in hyper spectral thermal measurements forms the foundation of the NUCAPS retrieval algorithm. Secondly, NUCAPS is a modular, name-list driven design. It can process multiple hyper spectral infrared sounders (on Aqua, NPP, MetOp and JPSS series) by mean of the same exact retrieval software executable and underlying spectroscopy. Finally, a cloud-clearing algorithm and a synergetic use of microwave radiance measurements enable full vertical sounding of the atmosphere, under all-sky regimes. As we transition toward improved hyper spectral missions, assessing retrieval skill and consistency across multiple platforms becomes a priority for real time users applications. Focus of this presentation is a general introduction on the recent improvements in the delivery of the NUCAPS full spectral resolution upgrade and an overview of the lessons learned from the 2017 Hazardous Weather Test bed Spring Experiment. Test cases will be shown on the use of NPP and Met

  4. Strengthening leadership as a catalyst for enhanced patient safety culture

    DEFF Research Database (Denmark)

    Kristensen, Solvejg; Christensen, Karl Bang; Jaquet, Annette

    2016-01-01

    OBJECTIVES: Current literature emphasises that clinical leaders are in a position to enable a culture of safety, and that the safety culture is a performance mediator with the potential to influence patient outcomes. This paper aims to investigate staff's perceptions of patient safety culture...... in a Danish psychiatric department before and after a leadership intervention. METHODS: A repeated cross-sectional experimental study by design was applied. In 2 surveys, healthcare staff were asked about their perceptions of the patient safety culture using the 7 patient safety culture dimensions...... in the Safety Attitudes Questionnaire. To broaden knowledge and strengthen leadership skills, a multicomponent programme consisting of academic input, exercises, reflections and discussions, networking, and action learning was implemented among the clinical area level leaders. RESULTS: In total, 358 and 325...

  5. Current status of NPP generation IV

    International Nuclear Information System (INIS)

    Yohanes Dwi Anggoro; Dharu Dewi; Nurlaila; Arief Tris Yuliyanto

    2013-01-01

    Today development of nuclear technology has reached the stage of research and development of Generation IV nuclear power plants (advanced reactor systems) which is an innovative development from the previous generation of nuclear power plants. There are six types of power generation IV reactors, namely: Very High Temperature Reactor (VHTR), Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), and Super Critical Water-cooled Reactor (SCWR). The purpose of this study is to know the development of Generation IV nuclear power plants that have been done by the thirteen countries that are members of the Gen IV International Forum (GIF). The method used is review study and refers to various studies related to the current status of research and development of generation IV nuclear power. The result of this study showed that the systems and technology on Generation IV nuclear power plants offer significant advances in sustainability, safety and reliability, economics, and proliferation resistance and physical protection. In addition, based on the research and development experience is estimated that: SFR can be used optimally in 2015, VHTR in 2020, while NPP types GFR, LFR, MSR, and SCWR in 2025. Utilization of NPP generation IV said to be optimal if fulfill the goal of NPP generation IV, such as: capable to generate energy sustainability and promote long-term availability of nuclear fuel, minimize nuclear waste and reduce the long term stewardship burden, has an advantage in the field of safety and reliability compared to the previous generation of NPP and VHTR technology have a good prospects in Indonesia. (author)

  6. PTS assessment - The basis of life time evaluation at NPP Paks

    International Nuclear Information System (INIS)

    Elter, J.; Oszwald, F.; Ratkay, S.; Fekete, T.; Gillemot, F.; Marothy, L.

    1997-01-01

    Plant specific PTS analysis at NPP Paks was performed in the frame of the AGNES (Advanced General New Evaluation of Safety) project. NPP Paks belongs to the second generation of the WWER-440/213 NPP-s. To verify the safety during transient events and predict the lifetime of the RPV-s several transient cases have been analyzed. The paper summarizes: The general scheme elaborated for the assessment; the safety philosophy used; the applied and available codes and methods; the ongoing and planned developments. (author). 8 refs, 3 figs, 1 tab

  7. PTS assessment - The basis of life time evaluation at NPP Paks

    Energy Technology Data Exchange (ETDEWEB)

    Elter, J; Oszwald, F; Ratkay, S [NPP Paks (Hungary); Fekete, T; Gillemot, F; Marothy, L [Atomic Energy Research Inst., Budapest (Hungary)

    1997-09-01

    Plant specific PTS analysis at NPP Paks was performed in the frame of the AGNES (Advanced General New Evaluation of Safety) project. NPP Paks belongs to the second generation of the WWER-440/213 NPP-s. To verify the safety during transient events and predict the lifetime of the RPV-s several transient cases have been analyzed. The paper summarizes: The general scheme elaborated for the assessment; the safety philosophy used; the applied and available codes and methods; the ongoing and planned developments. (author). 8 refs, 3 figs, 1 tab.

  8. Enhancing Public Helicopter Safety as a Component of Homeland Security

    Science.gov (United States)

    2016-12-01

    Risk Assessment Tool GPS Global Positioning System IFR instrument flight rules ILS instrument landing system IMC instrument meteorological...daily operations. Additionally, the effectiveness of the standards is evaluated by determining if these standards would have prevented the accidents...trends, such as human behavior and lack of standards, that are common in public safety helicopter accidents. Public safety aviation agencies can use this

  9. Enhancing vaccine safety capacity globally: A lifecycle perspective

    NARCIS (Netherlands)

    R.T. Chen (Robert T.); T.T. Shimabukuro (Tom T.); D.B. Martin (David); P. Zuber (Patrick); D.M. Weibel (Daniel); M.C.J.M. Sturkenboom (Miriam)

    2015-01-01

    textabstractMajor vaccine safety controversies have arisen in several countries beginning in the last decades of 20th century. Such periodic vaccine safety controversies are unlikely to go away in the near future as more national immunization programs mature with near elimination of target

  10. Contractors’ Attitude towards Enhancing Safety Performance: Case Study on Construction Firms in Penang

    Directory of Open Access Journals (Sweden)

    Ulang N. Md

    2014-01-01

    Full Text Available A qualitative study was conducted to investigate the contractors’ attitude towards enhancing the safety performance in construction site. Despite the fact that there are many safety initiatives established by the government, the rates of accidents are still in a critically high condition. Thus the purpose of this research is to study the contractors’ attitude towards enhancing the implementation of safety management system in construction site in order to increase the safety awareness of construction practitioners in construction site and improve the safety condition of construction sites. This study is conducted through oral interviews with the construction practitioners, and visual inspection of construction sites. The attitudes of contractors are evaluated from 3 aspects: Contractors’ efforts in implement and enforce the safety rules, Contractors efforts in overcoming the rate of accidents, and Reasons given by the contractors for not implement safety law.

  11. External Events PSA for the Paks NPP

    International Nuclear Information System (INIS)

    Bareith, Attila; Karsa, Zoltan; Siklossy, Tamas; Vida, Zoltan

    2014-01-01

    Initially, probabilistic safety assessment of external events was limited to the analysis of earthquakes for the Paks Nuclear Power Plant in Hungary. The level 1 seismic PSA was completed in 2002 showing a significant contribution of seismic failures to core damage risk. Although other external events of natural origin had previously been screened out from detailed plant PSA mostly on the basis of event frequencies, a review of recent experience on extreme weather phenomena made during the periodic safety review of the plant led to the initiation of PSA for external events other than earthquakes in 2009. In the meantime, the accident of the Fukushima Dai-ichi Nuclear Power Plant confirmed further the importance of such an analysis. The external event PSA for the Paks plant followed the commonly known steps: selection and screening of external hazards, hazard assessment for screened-in external events, analysis of plant response and fragility, PSA model development, and risk quantification and interpretation of results. As a result of event selection and screening the following weather related external hazards were subject to detailed analysis: extreme wind, extreme rainfall (precipitation), extreme snow, extremely high and extremely low temperatures, lightning, frost and ice formation. The analysis proved to be a significant challenge due to scarcity of data, lack of knowledge, as well as limitations of existing PSA methodologies. This paper presents an overview of the external events PSA performed for the Paks NPP. Important methodological aspects are summarised. Key analysis findings and unresolved issues that need further elaboration are highlighted. Development of external events PSA for the Paks NPP was completed by the end of 2012. The analysis followed the commonly known steps: selection and screening of external hazards, hazard assessment for screened-in external events, analysis of plant response and fragility, PSA model development, and risk

  12. Hydrodynamics and sediment transport at Muria Peninsula NPP Site

    International Nuclear Information System (INIS)

    Heni Susiati; Berni A Subki; Harman A

    2011-01-01

    Coastal along the coast of the Muria Peninsula, particularly the location of the Muria NPP site candidate is a dynamic region, the interaction between physical oceanographic factors such as currents, waves and tides in the coastal sediments cause abrasion or accretion. Interactions have resulted in coastal dynamics needs to be considered in siting NPP is essential in order to plan. Capacity of hydro-oceanographic data is essential in order to plan the development of the Muria NPP. The process of selecting a safe site for hydro-oceanographic aspects carried out according to IAEA safety standards on site selection. For the evaluation stage of hydro oceanographic potential site (site survey stage), the analysis is more focused on the tidal along the northern coast, bathymetry, potential water resources and hydrologic systems in the Muria NPP siting locations, Jepara. The method used is a secondary, confirmation of field data collection and interpretation of modeling results. The results showed that the preparation for the construction of NPP need to be evaluated further to coastal conditions with respect to the increase coastal erosion in the area of prospective NPP siting. (author)

  13. IRSN - Annual Report 2013. Financial Report 2013. Enhancing nuclear safety

    International Nuclear Information System (INIS)

    Schuler, Matthieu; Marchal, Valerie; Albert, Marc-Gerard; Aurelle, Jacques; Bigot, Marie-Pierre; Bruna, Giovanni; Charron, Sylvie; Clavelle, Stephanie; Cousinou, Patrick; Deschamps, Patrice; Delattre, Aleth; Demeillers, Didier; Dumas, Agnes; Franquard, Dominique; Laloi, Patrick; Lorthioir, Stephane; Monti, Pascale; Rollinger, Francois; Rouyer, Veronique; Rutschkovsky, Nathalie; Scott De Martinville, Edouard; Tharaud, Christine; Verpeaux, Jean-Luc; Jaunet, Camille; Hedouin, Jean-Christophe; Pascal-Heuze, Charlotte

    2014-03-01

    IRSN, a public entity with industrial and commercial activities, is placed under the joint authority of the Ministries of Defense, Environment, Industry, Research, and Health. It is the nation's public service expert in nuclear and radiation risks, and its activities cover all the related scientific and technical issues. Its areas of specialization include the environment and radiological emergency response, human radiation protection in both a medical and professional capacity, and in both normal and post-accident situations, the prevention of major accidents, nuclear reactor safety, as well as safety in nuclear plants and laboratories, transport and waste treatment, and nuclear defense and security expertise. IRSN interacts with all parties concerned by these risks (public authorities, in particular nuclear safety and security authorities, local authorities, companies, research organizations, stakeholders' associations, etc.) to contribute to public policy issues relating to nuclear safety, human and environmental protection against ionizing radiation, and the protection of nuclear materials, facilities, and transport against the risk of malicious acts. This document is the 2013 issue of IRSN's activity report. Content: 1 - Organization, key figures; 2 - Strategy: Progress and main activities in 2013, Transparency and communications policy, Promoting a safety and radiation protection culture; 3 - Activities: Safety (Safety of existing facilities, Studies and researches, About defense, Conducting assessments of future facilities); Nuclear security and non-proliferation (Nuclear security activities, International non-proliferation controls); Radiation protection - environment and human health (Environmental and population exposure, Radiation protection in the workplace, Effects of chronic exposure, Protection in health care); Emergency and post-accident situations efficiency; 4 - Efficiency: Health, safety, environmental, protection and quality, Human resources

  14. Safety of nuclear power plants in Slovak Republic in the context of integration into the European Union

    International Nuclear Information System (INIS)

    Pillar, V.

    2001-01-01

    A review of the construction and operation of nuclear power plants in Slovak Republic which belong within Slovenske Elektrarne is made. Since 1993 with the applying of the Least Cost Planning Methods and Financial Feasibility Analysis the optimal variant, considering construction completion of the two units at Mochovce NPP, has been adopted and a project for safety enhancement has been implemented. Some changes have been made in siting project in order to meet the seismic requirements, following by change of the control system design and supplier. The most important result of the safety enhancement program, important not only for NPP Mochovce, but also for all NPPs of WWER 440/213 type, is the fact, that operability of containment with passive bubbler-condenser system for pressure suppression and its acceptability from safety point of view, has been fully confirmed. Results of safety enhancement program have been re-evaluated not only by International Atomic Energy Agency, but also by other international authorities and professional organisations with positive result. Especially significant is the conclusion of RISKAUDIT, which has carried out repeated evaluation of safety level of NPP Mochovce based on PHARE project and which is accepted by European Commission. A similar approach for enhancing the safety has been applied for the NPP V-1 units at Jaslovske Bohunice

  15. The local impacts of Borssele NPP closing down

    International Nuclear Information System (INIS)

    Mandos, J.L.

    2000-01-01

    The municipal council of Borsele considers that the use of nuclear energy is a question of national policy, but considers itself responsible for town and country planning, safety and employment. The two latter aspects are especially significant if plans for shutdown are concerned. Confidence in safety of the NPP in Borsele has increased in recent years due to transparent information policy of the power plant itself as well as the local government and most of the employees living in the area. If considering the circumstances under which the decision about the closure of NPP one could sees this is a purely political decision. The local government will continue the close cooperation with the National government and the Parliament in solving the problem of employment and social impact of NPP shutdown

  16. NPP unusual events: data, analysis and application

    International Nuclear Information System (INIS)

    Tolstykh, V.

    1990-01-01

    Subject of the paper are the IAEA cooperative patterns of unusual events data treatment and utilization of the operating safety experience feedback. The Incident Reporting System (IRS) and the Analysis of Safety Significant Event Team (ASSET) are discussed. The IRS methodology in collection, handling, assessment and dissemination of data on NPP unusual events (deviations, incidents and accidents) occurring during operations, surveillance and maintenance is outlined by the reports gathering and issuing practice, the experts assessment procedures and the parameters of the system. After 7 years of existence the IAEA-IRS contains over 1000 reports and receives 1.5-4% of the total information on unusual events. The author considers the reports only as detailed technical 'records' of events requiring assessment. The ASSET approaches implying an in-depth occurrences analysis directed towards level-1 PSA utilization are commented on. The experts evaluated root causes for the reported events and some trends are presented. Generally, internal events due to unexpected paths of water in the nuclear installations, occurrences related to the integrity of the primary heat transport systems, events associated with the engineered safety systems and events involving human factor represent the large groups deserving close attention. Personal recommendations on how to use the events related information use for NPP safety improvement are given. 2 tabs (R.Ts)

  17. Role of field testing and shaking table test on full scale structure for NPP seismic-safety, and its relation to computational mechanics

    International Nuclear Information System (INIS)

    Shibata, Heki

    1988-01-01

    Field testing on the dynamic behavior of actual structures is significant for the seismic safety of nuclear power plants. For their mechanical components and piping systems, the full scale testings are also important as well as the in-situ test of buildings. In general, it is often observed that they don't behave as that of analytical model for the design. This article tries to discuss how such discrepancy is occurring, and how to overcome it. (author)

  18. Proceedings of a joint OECD/NEA-IAEA symposium on human factors and organisation in NPP maintenance outages: impact on safety

    International Nuclear Information System (INIS)

    1995-01-01

    The sessions of this conference dealt with outage strategy and methods (in Sweden, France and United States), the organisation and management of outages (organisation during refuelling shutdowns in France, safety approaches in France, in the USA, in Canada, in the United Kingdom and in Sweden), case studies and lessons learned (in France, Korea, Sweden, UK, USA), regulatory aspects of outages (UK, Germany, Mexico, France), the development of outage techniques

  19. Role of field testing and shaking table test on full scale structure for NPP seismic-safety, and its relation to computational mechanics

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Heki [Institute of Industrial Science, University of Tokyo (Japan)

    1988-07-01

    Field testing on the dynamic behavior of actual structures is significant for the seismic safety of nuclear power plants. For their mechanical components and piping systems, the full scale testings are also important as well as the in-situ test of buildings. In general, it is often observed that they don't behave as that of analytical model for the design. This article tries to discuss how such discrepancy is occurring, and how to overcome it. (author)

  20. Russian normative approach to the question of management of NPP life time

    International Nuclear Information System (INIS)

    Karpunin, N.I.

    2002-01-01

    Full text: In Russia, the designated service life of a nuclear power plant (NPP) is 30 years. During the period 2001-2010, 15 Russian NPP units will reach the end of their service life. The 'Basic Provisions of NPP Safety Assurance', OPB-88/97, Item 5.1.14, provide for a possible extension of NPP operation beyond the designated service life. For such an extension, the NPP operating organization must apply for a license renewal to Gosatomnadzor, which needs to specify the relevant requirements. GAN is developing regulatory documents to provide a basis for NPP license renewal/extension of NPP operation, which would benefit from international experience. In accordance with 'The program of Atomic Energy Development in the Russian Federation for 1998-2005 and up to 2010' adopted by Decree No. 815 of the Government of the Russian Federation on 21 July 1998, priority is placed on the preparation of NPPs for extension of service life and on ensuring safety in the extended operating period. The length of the extension beyond the designated service life is to be determined on the basis of a range of technical and economic considerations, including: The ability to ensure and maintain operational safety; Sufficient residual service life of the unit's non-repairable components; The availability of temporary storage facilities for the additional spent nuclear fuel and radioactive waste, or the possibility of its transport off-site; The ability to ensure safe handling of the radioactive waste generated during the extension period; To extend the lifetime of an NPP unit, the plant Operator is required to perform the following tasks: Carry out a comprehensive survey of the NPP unit; Draw up a programme of preparation for lifetime extension; Prepare the NPP unit for operation in the extended period; Carry out the necessary tests. There are also some normative documents, which regulate management of NPP life time. (author)

  1. Enhanced safety of radiation workers: a regulatory approach

    Energy Technology Data Exchange (ETDEWEB)

    Gopalakrishnan, A [Atomic Energy Regulatory Board, Bombay (India)

    1994-04-01

    Radiation safety should not only be strictly implemented, but also believed and understood by the workers, the unions, the media and the general public as being fairly and adequately enforced. It is not at all sufficient that only those in the operational management levels satisfy themselves that workers` safety is properly taken care of, but it is necessary that the workers and their unions are also convinced about it and share this management view.

  2. Enhanced safety of radiation workers: a regulatory approach

    International Nuclear Information System (INIS)

    Gopalakrishnan, A.

    1994-01-01

    Radiation safety should not only be strictly implemented, but also believed and understood by the workers, the unions, the media and the general public as being fairly and adequately enforced. It is not at all sufficient that only those in the operational management levels satisfy themselves that workers' safety is properly taken care of, but it is necessary that the workers and their unions are also convinced about it and share this management view

  3. Plant performance monitoring program at Krsko NPP

    International Nuclear Information System (INIS)

    Bach, B.; Kavsek, D.

    2004-01-01

    A high level of nuclear safety and plant reliability results from the complex interaction of a good design, operational safety and human performance. This is the reason for establishing a set of operational plant safety performance indicators, to enable monitoring of both plant performance and progress. Performance indicators are also used for setting challenging targets and goals for improvement, to gain additional perspective on performance relative to other plants and to provide an indication of a potential need to adjust priorities and resources to achieve improved overall plant performance. A specific indicator trend over a certain period can provide an early warning to plant management to evaluate the causes behind the observed changes. In addition to monitoring the changes and trends, it is also necessary to compare the indicators with identified targets and goals to evaluate performance strengths and weaknesses. Plant Performance Monitoring Program at Krsko NPP defines and ensures consistent collection, processing, analysis and use of predefined relevant plant operational data, providing a quantitative indication of nuclear power plant performance. When the program was developed, the conceptual framework described in IAEA TECDOC-1141 Operational Safety Performance Indicators for Nuclear Power Plants was used as its basis in order to secure that a reasonable set of quantitative indications of operational safety performance would be established. Safe, conservative, cautious and reliable operation of the Krsko NPP is a common goal for all plant personnel. It is provided by continuous assurance of both health and safety of the public and employees according to the plant policy stated in program MD-1 Notranje usmeritve in cilji NEK, which is the top plant program. Establishing a program of monitoring and assessing operational plant safety performance indicators represents effective safety culture of plant personnel.(author)

  4. A proposed approach for enhancing design safety assurance of future plants

    International Nuclear Information System (INIS)

    Oh, Kyu Myeng; Ahn, Sang Kyu; Lee, Chang Ju; Kim, Inn Seock

    2010-01-01

    This paper provides various insights from a detailed review of deterministic approaches typically applied to ensure design safety of nuclear power plants (NPPs) and risk-informed approaches proposed to evaluate safety of advanced reactors such as Generation IV reactors. Also considered herein are the risk-informed safety analysis (RISA) methodology suggested by Westinghouse as a means to improve the conventional accident analysis, together with the Technology Neutral Framework recently suggested by the U.S. NRC for safety evaluation of future plants. These insights from the comparative review of deterministic and risk-informed approaches could be used in further enhancing the methodology for design safety assurance of future plants

  5. Preparation for Ignalina NPP decommissioning

    International Nuclear Information System (INIS)

    Medeliene, D.

    2004-01-01

    Latest developments of atomic energy in Lithuania, works done to prepare Ignalina NPP for final shutdown and decommissioning are described. Information on decommissioning program for Ignalina NPP unit 1, decommissioning method, stages and funding is presented. Other topics: radiation protection, radioactive waste management and disposal. Key facts related to nuclear energy in Lithuania are listed

  6. Current technological trends in development of NPP systems

    International Nuclear Information System (INIS)

    Florescu, Gheorghe; Panaitescu, Valeriu

    2010-01-01

    The recent nuclear research issues look for new technologies and continuous progress in finding different and efficient solutions for sustained and upraising energy demand. The trend of increasing energy consumption and occurring of new and large consumers, especially from Asian countries, imposes finding of new means for clean, large scale and sustained energy production. NPPs availability was continuously monitored and improved; at the same time the safety of the nuclear energy production was under surveillance. The present development of the new technologies, the discoveries of new materials and development of efficient technological processes offer the opportunities for their appropriate implementation and use in the NPP system configurations and functioning/operation. The new technologies and scientific discoveries, and also the international cooperation, offer the opportunities to mitigate the actual barriers in order to cumulate and use advanced energy production, to find new energy sources and to build improved, reliable and safe power plants. The monitoring systems, intelligent sensors and SSCs, nanotechnologies and new/intelligent materials constitute the main ways for improvement of the NPP systems configuration and processes. The paper presents: - The state of the art in the level of the currently applied technologies for nuclear power systems development; - The actual technological limits that need to be over passed for improving the NPP systems ; - The main systems that need improvement and reconfiguration for development of currently operating NPPs as well as raising the operation efficiency, availability and total safety; - The actual energy production issues; - The key arguments in sustaining the R and D new NPP systems development; - Future trends in NPP development; - The limitations in industrial processes knowledge and use. Appropriate R and D in the field of NPP systems have specific characteristics that were considered in paper completion

  7. Optimization of radiation protection at Bohunice NPP

    International Nuclear Information System (INIS)

    Dobis, L.; Svitek, J.

    2003-01-01

    Bohunice Nuclear Power Plant is situated in south - western part of Slovakia about 50 km away from Bratislava. There are four PWR reactors 440 MW e each - two units with reactors WWER - 230 (V1 NPP) and two units with WWER - 213 (V2 NPP). requirements for the optimization process are given in the mentioned Code No.12 of Ministry of Health. Code 12 stipulates the technical and organizational requirements for proving the Rational Achievable Level (RAL) of radiation protection. This level can be proved by means of the comparison of the dose distribution to the costs of protection. An example of two figures of dose constraints is: collective dose 20 man mSv for the specific task; individual exposure 1 mSv per day. The values of the financial equivalents of personal exposure - so called the alpha coefficients - are used for the calculation of the benefit of proposed measures. Impact of legislative changes into Bohunice NPP and optimization process are presented. Apparently the new law and the associate code created a base of transparent and understandable policy of radiation protection and optimization in Slovak Republic. The radiation protection legislative was implemented into the praxis and persons became familiar with it. Defining clear and unambiguous terms facilitated the communication between users and the regulatory body - State Health Institute. Optimization was generally accepted by the workers and managers and began to be a part of safety culture of operation at nuclear power plants. (authors)

  8. NPP Krsko on-line low pressure containment tightness monitoring implementation

    International Nuclear Information System (INIS)

    Dudas, M.; Basic, I.

    2004-01-01

    Containment Integrated Leak Rate Test (CILRT) 1999 in NPP Krsko was completely performed following regulation of 10CFR50 Appendix J Option A and ANSI/ANS 56.8-1987 at a design pressure (3.15 kp/cm2). In 2001 NPP Krsko proposed to Slovenian Nuclear Safety Administration (SNSA) the Technical Specification (TS) and Updated Safety Analysis Report (USAR) changes that describe implementation of new test intervals for Type A, B and C tests according to 10CFR50, Appendix J, Option B. After the positive final independent review of proposed changes by Authorized Institution, NPP Krsko received the License Amendment requiring from NPP Krsko to define technical solution for surveillance of containment tightness between two 10-years CILRT. This paper intends to discuss proposed methods by NPP Krsko, test equipment, performed measurements in 2004, associated analyses and evaluation.(author)

  9. 'HSE protection through safe work'. Mochovce NPP is the holder of ISO 14001:2004 and CSN OHSAS 18001 certificates

    International Nuclear Information System (INIS)

    Molnarova, Zuzana; Torda, Anton

    2009-01-01

    The article provides information on the safety of the Mochovce NPP related to occupational health and safety and environmental protection. The facility has earned EN ISO 140001 and OHSAS 18001 certificates. (orig.)

  10. Simulation research to enhance patient safety and outcomes: recommendations of the Simnovate Patient Safety Domain Group

    OpenAIRE

    Pucher, PH; Tamblyn, R; Boorman, D; Dixon-Woods, Mary Margaret; Donaldson, L; Draycott, T; Forster, A; Nadkarni, V; Power, C; Sevdalis, N; Aggarwal, R

    2017-01-01

    The use of simulation-based training has established itself in healthcare but its implementation has been varied and mostly limited to technical and non-technical skills training. This article discusses the possibilities of the use of simulation as part of an overarching approach to improving patient safety, and represents the views of the Simnovate Patient Safety Domain Group, an international multidisciplinary expert group dedicated to the improvement of patient safety. The application and ...

  11. Enhancing the Safety Climate and Reducing Violence Against Staff in Closed Hospital Wards.

    Science.gov (United States)

    Isaak, Valerie; Vashdi, Dana; Bar-Noy, Dor; Kostisky, Hava; Hirschmann, Shmuel; Grinshpoon, Alexander

    2017-09-01

    This study examined the effectiveness of an intervention program to enhance unit safety climate and minimize employee risk of injury from patient violence. The intervention program, including a 3-day workshop, was offered to personnel on maximum security units of an Israeli psychiatric hospital. Safety climate was examined before and after the implementation of the intervention, and incidents of patient violence were investigated. Six months after the intervention, a significant improvement in employees' perceptions of management's commitment to safety as well as a marginally significant improvement in communication about safety issues were found. This study demonstrated that an intervention program to enhance safety climate was associated with a decrease in the number of aggressive incidents. The researchers concluded that this intervention program is likely to return a sense of safety to workers and reduce workplace violence.

  12. Perspectives of Living PSA in NPP Krsko

    International Nuclear Information System (INIS)

    Vrbanic, I.; Kastelan, M.

    1996-01-01

    Nuclear power plant Krsko has completed the Level 1/Level 2 Probabilistic Safety Analysis (PSA) for internal initiating events and is in the process of completing the same for the external initiators. The analysis completed up to now has provided a valuable insight into a plant risk profile. In NPP Krsko there is a plan to use the PSA model as a permanent tool for the risk based applications and incorporate it into a decision making process. In order to achieve this there is a need to permanently maintain the PSA model in a manner that it reflects both the plan configuration/design at a time point and the operational experience up to the time point. All the activities aimed toward keeping the PSA model up-to-dated in this sense are usually referred to as a Living PSA (LPSA) program. NPP Krsko is in the process of defining and proceduralizing a LPSA program that would be plant specific and based on known world practices. Further, in order to be suitable for risk based applications the PSA model must be flexible in a sense that modifications to the base case model may be done easily and requantifications performed quickly as to evaluate various conditions imposed by real or hypothetical situations. NPP Krsko PSA model has been based on licensing type software. The requirements specified above dictate the transfer of the overall model to an application oriented software of newer generation with larger capabilities. The transfer becomes a part of a mentioned ongoing effort aimed at establishing LPSA model and concept. The paper present this effort and the perspectives of LPSA concept and risk based applications in NPP Krsko. (author)

  13. An approach to Identify the Risk Induced by Cyber-Attack on the Non-safety NPP I and C System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Eun; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Kim, Jong Hyun [Chosun University, Gwangju (Korea, Republic of); Son, Han Sung [Joonbu University, Geumsan (Korea, Republic of)

    2016-05-15

    In this study, influence of the attack on the non-safety system will be investigated, because the cyber-attack on the safety system cannot be accomplished easily. To identify the risk from cyber-attack, the result of PSA will be applied. Cyber-attack may cause other risks except for the core damage. Those risks also can be identified by applying this method. This study could be reinforced in a more realistic way if the information on the maintenance is considered, because certain type of cyber-attack could be detected during the maintenance. Also, possible set of wrong actions need to be selected, based on the knowledge of I and C system and its vulnerabilities because the hacker might not attack every information. To obtain the realistic result information that can be manipulated need to be listed, because the hacker may not attack certain information, not to be detected during the maintenance. In addition, by using the result of this study, the test plan for the cyber-attack can be suggested. If the scenario is given, the criteria for the test target selection can be obtained. It includes the target component and information.

  14. An approach to Identify the Risk Induced by Cyber-Attack on the Non-safety NPP I and C System

    International Nuclear Information System (INIS)

    Kim, Hee Eun; Kang, Hyun Gook; Kim, Jong Hyun; Son, Han Sung

    2016-01-01

    In this study, influence of the attack on the non-safety system will be investigated, because the cyber-attack on the safety system cannot be accomplished easily. To identify the risk from cyber-attack, the result of PSA will be applied. Cyber-attack may cause other risks except for the core damage. Those risks also can be identified by applying this method. This study could be reinforced in a more realistic way if the information on the maintenance is considered, because certain type of cyber-attack could be detected during the maintenance. Also, possible set of wrong actions need to be selected, based on the knowledge of I and C system and its vulnerabilities because the hacker might not attack every information. To obtain the realistic result information that can be manipulated need to be listed, because the hacker may not attack certain information, not to be detected during the maintenance. In addition, by using the result of this study, the test plan for the cyber-attack can be suggested. If the scenario is given, the criteria for the test target selection can be obtained. It includes the target component and information

  15. Structural safety test and analysis of type IP-2 transport packages with bolted lid type and thick steel plate for radioactive waste drums in a NPP

    International Nuclear Information System (INIS)

    Kim, Dong Hak; Seo, Ki Seog; Lee, Sang Jin; Lee, Kyung Ho; Kim, Jeong Mook

    2007-01-01

    If a type IP-2 transport package were to be subjected to a free drop test and a penetration test under the normal conditions of transport, it should prevent a loss or dispersal of the radioactive contents and a more than 20% increase in the maximum radiation level at any external surface of the package. In this paper, we suggested the analytic method to evaluate the structural safety of a type IP-2 transport package using a thick steel plate for a structure part and a bolt for tying a bolt. Using an analysis a loss or disposal of the radioactive contents and a loss of shielding integrity were confirmed for two kinds of type IP-2 transport packages to transport radioactive waste drums from a waste facility to a temporary storage site in a nuclear power plant. Under the free drop condition the maximum average stress at the bolts and the maximum opening displacement of a lid were compared with the tensile stress of a bolt and the steps in a lid, which were made to avoid a streaming radiation in the shielding path, to evaluate a loss or dispersal of radioactive waste contents. Also a loss of shielding integrity was evaluated using the maximum decrease in a shielding thickness. To verify the impact dynamic analysis for free drop test condition and evaluate experimentally the safety of two kinds of type IP-2 transport packages, free drop tests were conducted with various drop directions

  16. ENHANCING FOOD SAFETY AND STABILITY THROUGH IRRADIATION: A REVIEW

    Directory of Open Access Journals (Sweden)

    Manzoor Ahmad Shah

    2014-04-01

    Full Text Available Food irradiation is one of the non thermal food processing methods. It is the process of exposing food materials to the controlled amounts of ionizing radiations such as gamma rays, X-rays and accelerated electrons, to improve microbiological safety and stability. Irradiation disrupts the biological processes that lead to decay of food quality. It is an effective tool to reduce food-borne pathogens, spoilage microorganisms and parasites; to extend shelf-life and for insect disinfection. The safety and consumption of irradiated foods have been extensively studied at national levels and in international cooperations and have concluded that foods irradiated under appropriate technologies are both safe and nutritionally adequate. Specific applications of food irradiation have been approved by national legislations of more than 55 countries worldwide. This review aims to discuss the applications of irradiation in food processing with the emphasis on food safety and stability.

  17. Enhancing nuclear safety. Annual report 2014. Financial report 2014

    International Nuclear Information System (INIS)

    2015-01-01

    After some introductory texts proposed by several IRSN head managers, and a brief presentation of some key data illustrating the activity, the annual report presents the main strategic orientations, notably in the field of knowledge management, and of information and communication. After some images illustrating the past year, activities are presented. They first deal with safety: Reactor safety (operating experience feedback), From decommissioning old reactors to designing those of the future, Safety of laboratories and plants, Safety regarding risks due to infrastructure near nuclear facilities, Reactor aging, Fuel: research on corrosion and deformation, Research and assessments for improved understanding of accident situations, Earthquakes: research and assessments, About defense, Geological disposal of radioactive waste. They secondly deal with security and non-proliferation (nuclear security, nuclear non-proliferation, chemical weapon ban), thirdly with radiation protection for human and environment health (environment monitoring, radionuclide transfer in the environment, radon and polluted sites, human exposure, radiation protection in the workplace, effects of low-dose chronic exposures, Organization of radiation protection at the European level, protection in health care), and fourthly with emergency and post-accident situations (emergency and post-accident preparedness and response, Emergency response tools). The next part of the activity report addresses issues related to efficiency: Real estate program (construction projects get started), Hygiene, safety, social responsibility, Human resources, Organization chart, Board of directors, Steering committee for the nuclear defense expertise Division - CODEND, Scientific council, Ethics commission composition, Nuclear safety and radiation protection Research policy committee - COR. The financial report proposes a management report, financial statements with an appendix to annual accounts, and an auditor

  18. Enhanced FAA-hybrid III numerical dummy model in Madymo for aircraft occupant safety assessment

    NARCIS (Netherlands)

    Boucher, H.; Waagmeester, C.D.

    2003-01-01

    To improve survivability and to minimize the risk of injury to occupants in helicopter crash events, a complete Cabin Safety System concept including safety features and an enhanced FAA-Hybrid III dummy were developed within the HeliSafe project. A numerical tool was also created and validated to

  19. Accidents during shutdown conditions for NPP with WWER-1000/428

    International Nuclear Information System (INIS)

    Antropov, H.A.

    1996-01-01

    This paper considers the effect of such internal initiating events upon safety of the NPP with WWER-1000/428 as boron dilution and loss of residual heat removal at the reduced coolant inventory in the primary circuit. 2 refs

  20. The problem of gas gap between graphite - fuel channel reduction impact at Ignalina NPP

    International Nuclear Information System (INIS)

    1999-01-01

    Safety analysis of Ignalina NPP operation in the case when gap closure between graphite - fuel channel occur was performed. The main results of this analysis as well as data of gap measurements during the year 1996 - 1998 are provided

  1. Comparative analysis of events at NPP of Ukraine during 1993-1995

    Energy Technology Data Exchange (ETDEWEB)

    Koltakov, V [Ministry for Environment Protection and Nuclear and Radiation Safety of Ukraine Nuclear Regulatory Administration Scientific and Technical Centre, Kiev (Ukraine)

    1997-12-31

    The suggested comparative analyses of malfunctions in operation of Ukrainian NPP during the last 3 year shows a tendency for decreasing the total number of malfunctions, out of plan power changes and malefactions connected with safety system nonavailability. 6 tabs.

  2. Comparative analysis of events at NPP of Ukraine during 1993-1995

    International Nuclear Information System (INIS)

    Koltakov, V.

    1996-01-01

    The suggested comparative analyses of malfunctions in operation of Ukrainian NPP during the last 3 year shows a tendency for decreasing the total number of malfunctions, out of plan power changes and malefactions connected with safety system nonavailability. 6 tabs

  3. Analysis of flooding possibilities of premises of Ignalina NPP

    International Nuclear Information System (INIS)

    1999-01-01

    Analysis of flooding possibilities of premises with safety related equipment of Ignalina NPP was performed. In calculations codes SQUIRT and RELAP5 was used. In most cases it was found that it would be enough time to take corrective measure in the case of flooding accident

  4. Challenges and Enhancements to the Safety Culture of the Regulatory Body

    International Nuclear Information System (INIS)

    Niel, Jean-Christophe; Chevet, Pierre Franck; Sheron, Brian; Boyd, Michael; Carlsson, Lennart; Tiippana, Petteri; Burns, Stephen; Jamieson, Terry; Fuketa, Toyoshi; Rzentkowski, Greg; Weiss, Frank Peter; Le Guen, Bernard

    2015-06-01

    The workshop opened with presentations by both the NEA Director-General and the chair of the three committees directly involved with the safety culture of the regulatory body (SCRB). The opening session set the scene and gave an overview of the SCRB together with presentations and discussions on priorities and challenges. The main session focused on the principles of the SCRB, its implementation and the challenges and enhancements that are being raised and considered. The workshop concluded with a session that looked at findings and conclusions, the way forward and an agreed position on the SCRB. This document brings together the available presentations (slides) given at the workshop: 1 - Introduction: Challenges and Enhancements to the Safety Culture of the Regulatory Body (J-C. Niel); 2 - Thoughts on Safety Culture from a CSNI Perspective (B. Sheron); 3 - Radiological Protection Culture: CRPPH Work (M. Boyd); 4 - Challenges and Enhancements to Safety Culture of the Regulatory Body (L. Carlsson); 5 - Principles for the safety culture of the regulatory body (P. Tiippana); 6 - NRC's Internal Safety Culture: Successes, Challenges, and the Path Forward (S.G. Burns); 7 - Insights on the Canadian Nuclear Safety Commission's Safety Culture Journey (T. Jamieson); 8 - Lessons Learned from the Fukushima Dai-ichi Accident regarding Safety Culture of Regulatory Body (T. Fuketa); 9 - Challenges to Regulatory Bodies' safety culture (P-F. Chevet); 10 - Regulatory Safety Culture: International Perspective (G. Rzentkowski); 11 - Integration of Safety Research into Safety Culture Concepts (F-P. Weiss); 12 - Radiation Protection and Emergency Management Aspects: Culture drawn up by RP professionals (B. Le Guen); 13 - Closing session panel (L. Carlsson)

  5. Regional aviation safety organisations : enhancing air transport safety through regional cooperation

    NARCIS (Netherlands)

    Ratajczyk, Mikołaj Andrzej

    2014-01-01

    Aviation safety is of global concern. Yet 43% of countries worldwide lack the expertise and administrative capacity to ensure effective safety oversight of the aviation activities for which they are responsible under international law. With air traffic predicted to double in the next 15 years and

  6. WWER-type NPP spray ponds screen

    International Nuclear Information System (INIS)

    Nikolova, M.; Jordanov, M.; Denev, J.; Markov, D.

    2003-01-01

    The objective of this study is to develop a protection screen of WWER-type NPP spray ponds. The screen design is to ensure reduction of the water droplets blown by the wind and, if possible, their return back to the spray ponds. The cooling capacity of the ponds is not to be changed below the design level for safety reasons. Computational fluid dynamics analysis is used to assess the influence of each design variant on the behavior of the water droplets distribution. Two variants are presented here. The one with plants is found not feasible. The second variant, with steel screen and terrain profile modification is selected for implementation. (author)

  7. The accidents during shutdown conditions Temelin NPP

    International Nuclear Information System (INIS)

    Sykora, M.; Mlady, O.

    1996-01-01

    Two parallel activities oriented for the accidents during shutdown conditions are performed at Temelin NPP: Development of symptom based emergency operating procedures (EOPs) applicable for the accidents which could occur during operational modes 1 through 4; independent evaluation of plant safety as part of the Temelin Shutdown probabilistic assessment to define the accidents which could occur during mode 5 and 6 for which the EOPs must be extended. Both these activities are in progress now because Temelin plant is still in the construction phase

  8. Safe 15 Terawatt of Temelin NPP

    International Nuclear Information System (INIS)

    Sula, M.

    2010-01-01

    In this work author presents a project Safe 15 Terawatt realised on the Temelin NPP. This project is one of the eight key projects of the CEZ group, associated in the 'Programme of efficiency'. The project started in June 2007 with long-term goals for horizon of year 2012. The safety indicators will be reached of the first quarter level of world's nuclear power plant - by the end of the first decade. By the end of year 2012 we will have achieved annual production of 15 billion kWh - in the Czech Republic: 15 Terawatt.

  9. Core power capability verification for PWR NPP

    International Nuclear Information System (INIS)

    Xian Chunyu; Liu Changwen; Zhang Hong; Liang Wei

    2002-01-01

    The Principle and methodology of pressurized water reactor nuclear power plant core power capability verification for reload are introduced. The radial and axial power distributions of normal operation (category I or condition I) and abnormal operation (category II or condition II) are simulated by using neutronics calculation code. The linear power density margin and DNBR margin for both categories, which reflect core safety, are analyzed from the point view of reactor physics and T/H, and thus category I operating domain and category II protection set point are verified. Besides, the verification results of reference NPP are also given

  10. Conceptual studies of construction and safety enhancement of ocean SMART mounted on GBS

    International Nuclear Information System (INIS)

    Kim, Min-Gil; Lee, Kang-Heon; Kim, Seong Gu; Woo, Il-Guk; Han, Jeong-Hoon; Lee, Phill-Seung; Lee, Jeong Ik

    2014-01-01

    Highlights: • We suggested the concept of coupling the SMART to the GBS, and we made suggested improvements. • We describe the design concepts and GA of SMART ONPP. • We analyzed seismic feature of SMART ONPP preliminarily. • We suggested the concept of coupling the IPSS to the SMART ONPP, and we made suggested improvements. - Abstract: From the Fukushima accident, protection of NPPs from any imaginable natural disasters became very important. In this study, the authors suggest a new concept of ocean nuclear power plant (ONPP) by using SMART as a reference reactor, which is the most recent Small Modular Reactor (SMR) developed by Korea, to demonstrate that the proposed concept can improve the safety of NPP from earthquake and tsunami. The proposed concept utilizes Gravity Based Structure (GBS), which is a widely spread construction technique of offshore plants. Because, floating type or submerged type NPPs can be easily affected by severe ocean environments such as tsunamis and storms, additional safety features have to be added to the existing land based plant. In contrast, the newly proposed GBS-type ONPP does not require going through significant design modifications due to inherent characteristics of the construction method. The authors have demonstrated this concept can be applied to the large nuclear power plant in the previous work and will expand this concept for SMRs in this paper. The authors discuss the new concept by presenting design parameters, design requirements, and the new total general arrangement. Furthermore, due to the unique configuration of ONPP SMART, innovative passive safety features can be added to the existing SMART design. The performance of proposed concept to resist earthquake as well as newly added passive safety feature will be discussed by presenting simplified analysis results

  11. Conceptual studies of construction and safety enhancement of ocean SMART mounted on GBS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min-Gil, E-mail: gggggtt@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Kang-Heon, E-mail: welcome@kaist.ac.kr [Division of Ocean Systems Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Kim, Seong Gu, E-mail: skim07@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Woo, Il-Guk, E-mail: igwoo@dsme.co.kr [Department of Energy System R and D (Plant R and D), Daewoo Shipbuilding and Marine Engineering Co., Ltd., 221-17, Nonhyun-Dong, Gangnam-Gu, Seoul 135-010 (Korea, Republic of); Han, Jeong-Hoon, E-mail: jhhan1@dsme.co.kr [Department of Energy System R and D (Plant R and D), Daewoo Shipbuilding and Marine Engineering Co., Ltd., 221-17, Nonhyun-Dong, Gangnam-Gu, Seoul 135-010 (Korea, Republic of); Lee, Phill-Seung, E-mail: philseung@kaist.edu [Division of Ocean Systems Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2014-10-15

    Highlights: • We suggested the concept of coupling the SMART to the GBS, and we made suggested improvements. • We describe the design concepts and GA of SMART ONPP. • We analyzed seismic feature of SMART ONPP preliminarily. • We suggested the concept of coupling the IPSS to the SMART ONPP, and we made suggested improvements. - Abstract: From the Fukushima accident, protection of NPPs from any imaginable natural disasters became very important. In this study, the authors suggest a new concept of ocean nuclear power plant (ONPP) by using SMART as a reference reactor, which is the most recent Small Modular Reactor (SMR) developed by Korea, to demonstrate that the proposed concept can improve the safety of NPP from earthquake and tsunami. The proposed concept utilizes Gravity Based Structure (GBS), which is a widely spread construction technique of offshore plants. Because, floating type or submerged type NPPs can be easily affected by severe ocean environments such as tsunamis and storms, additional safety features have to be added to the existing land based plant. In contrast, the newly proposed GBS-type ONPP does not require going through significant design modifications due to inherent characteristics of the construction method. The authors have demonstrated this concept can be applied to the large nuclear power plant in the previous work and will expand this concept for SMRs in this paper. The authors discuss the new concept by presenting design parameters, design requirements, and the new total general arrangement. Furthermore, due to the unique configuration of ONPP SMART, innovative passive safety features can be added to the existing SMART design. The performance of proposed concept to resist earthquake as well as newly added passive safety feature will be discussed by presenting simplified analysis results.

  12. Mobilizing local safety nets for enhanced adaptive capacity to ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    22 avr. 2016 ... In Zimbabwe, the increased frequency of drought, flash floods, and unpredictable rainfall has added to recurrent food deficits for poor households who depend on rainfed farming on nutrient poor soils. This brief explores how the erosion of Zunde raMambo — a traditional community safety net mechanism ...

  13. Development of Safety Culture Indicators for HANARO

    International Nuclear Information System (INIS)

    Wu, Jong-Sup; Lee, Kye-Hong

    2007-01-01

    Safety culture is more important than a technical matter for the management of nuclear facilities. Some of the accidents that have occurred recently in nuclear plants are important as a social problem besides a technical problem. That's why the management of nuclear plants has been focused on the safety culture to improve confidence of nuclear facilities. As for a safety culture, there are difficulties in that a tangible result does not come out clearly in spite of an effort for a long time. Some IAEA guides and reports about a safety culture and its evaluation method for nuclear power plants (NPP) were published after the Chernobyl accident. Until now there is no tool to evaluate a safety culture of for research reactors. HANARO developed its own safety culture indicators based on the IAEA's documents. The purpose of the development of the safety culture indicators is to evaluate and enhance the safety attitude in HANARO

  14. Basic fracture toughness requirements for ferritic materials of nuclear class pressure retaining equipment in NPP

    International Nuclear Information System (INIS)

    Ning Dong; Yao Weida

    2005-01-01

    In this paper, theory basis on cold brittleness and anti-brittle fracture design of ferritic materials are introduced summarily and fracture toughness requirements for ferritic materials in ASME code for nuclear safety class pressure retaining equipment in NPP are summarized and evaluated. The results show that notch impact toughness requirements for materials relate to nuclear safety class of materials so as to ensure that brittle fracture of retaining pressure boundary in NPP can not occur. (authors)

  15. Control system for NPP powerfull turbines

    International Nuclear Information System (INIS)

    Osipenko, V.D.; Rozhanskij, V.E.; Rokhlenko, V.Yu.

    1985-01-01

    A control system for NPP 1000 MW turbines safety is described. The turbine safety system has a hydraulic drive to actuate in case of increasipg of rotational speed of a turbine rotor and an electrohydraulic drce to operate in case of pressure reduction in the lubrication system, axial displacement deviation, etc. The system is highly reliable due to application of a safety system without slide valves and long-term operation of hydraulic controls in guarding conditions; the system epsures multifunctional control with high accuracy and speed due to application of the intricate electronic part, high speed of response with a limited use of high pressure oil due to application of two-pressure pumps, pneumohydraulic accumulators and oil discharge valves. Steady-state serviceability of the system is maintained by devices for valve cooling dawn. A shockless change from electrohydraulic to hydraulic control channels is provided

  16. Principles of Defense-in-depth philosophy applied in NPP engineering management

    International Nuclear Information System (INIS)

    Wu Guangwei

    2011-01-01

    Based on the Defense-in-depth Concept in nuclear and radiation safety, Defense-in-depth Concept for design management of Nuclear Power Plant (NPP) is developed in this paper to analyze the feasibility and importance of the application of the basic principle: Defense-in-depth concept in NPP systems performed during the design control of NPP. This paper focuses on the NPP engineering management process, and according to the analysis of such process, 5 principles of Defense-in-depth Concept applied in NPP design management are raised: (1) preventing the non-conformities of design via effective design quality management system; (2) discovering and correcting non-conformities of design quality in time via design checkup and design review meeting; (3) carrying out timely analysis and treatment against design non-conformities which have been transferred to construction phase; (4) Assessing and judging the severe non-conformities in construction phase, putting forward treatment opinions and remedies accordingly so as to avoid the existence of such non-conformities in physical construction of NPP; (5) Paying 'return-visit' and performing 'post-assessment' for NPP design to assess the designed functions and safety of NPP comprehensively. (author)

  17. Probabilistic safety analysis of the Kozloduy NPP units 1-4 (WWER-440/230) using independent emergency feedwater system; Veroyatnostnyj analiz bezopasnosti I-IV blokov AEhS `Kozloduy` s reaktorami tipa WWER-440 (V 230) pri vklyuchenii nezavisimoj sistemy avarijnoj podpitki PG

    Energy Technology Data Exchange (ETDEWEB)

    Kalchev, B; Marinov, M; Dimitrov, B; Avdzhiev, K [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    The safety of the Kozloduy NPP is being promoted by backfitting and improved operational practice. Special measures mitigating potential severe accidents consequences are needed because of some deficiencies in the original design of the four WWER-440 units. In conditions of a total LOCA (Loss Of Coolant Accident) it is impossible to ensure decay heat removal using the existing safety system. In such cases an extra emergency feedwater system independent of the plant`s other systems has been introduced which offers a new alternative means of removing the residual heat from the reactor. A probabilistic safety analysis is carried out using the method of event trees. A comparison between the existing safety system and the newly proposed is made. The simulation results of the unit behaviour prove that the damage frequency of the active zone is lower with the new system. 3 refs., 3 tabs., 2 figs.

  18. Structure study and design of Qinshan NPP PCCV

    International Nuclear Information System (INIS)

    Xia Zufeng; Xu Yongzhi; Wang Tianzhen; Wu Jibiao

    1993-02-01

    The design process of Qinshan NPP (nuclear power plant) PCCV (prestressed concrete containment vessel) is summarized. The tendon test, structural description, design bases and analysis method are introduced. The arrangement for preventing concrete from cracking and design features of post-tensioning system and steel liner are presented. The results of model test and non-linear analysis for ultimate load in Qinshan NPP PCCV are also given. Through the integrity test of PCCV, it shows that the test values are in agreement with predicted values, the structure is excellent and the performance of leak tightness conforms to the safety requirements

  19. Study on the risk-informed regulation of NPP

    International Nuclear Information System (INIS)

    Wang Chaogui

    2007-01-01

    The risk-informed regulation is a modern type of NPP safety management mode using both deterministic and probabilistic approaches. It is necessary to entirely and systematically study the associated regulations, standards and practices in order to promote the developments of risk-informed regulations in China. This paper introduces the risk-informed regulation, gives out the basic principles, method and acceptance risk criteria of risk-informed decision,making, discusses the PSA requirements for risk-informed decision-making and makes some suggestions about the application of risk-informed regulations in Chinese NPP. (authors)

  20. Enhancing the Safety and Security of Radioactive Sources

    International Nuclear Information System (INIS)

    Hickey, J.

    2004-01-01

    The NRC initiatives to improve safety and security of sources began before 091101 and include both international and domestic activities. They supported the development and implementation of the IAEA Code of Conduct, which provides categorization of sources of concern, based on risk, improvement of regulatory programs of all member countries and improvement of safety and security of sources. International activities include the IAEA International Conference on Security of Sources (Vienna, Austria, March, 2003), the trilateral cooperation with Canada and Mexico, the assistance to individual countries to improve security and the proposed rule on export and import of radioactive material. The domestic initiatives are to issue the security orders and advisories to licensees, issue the panoramic irradiator orders (June 2003), issue the manufacturer orders (January 2004), complete the interim national source inventory, develop the national source tracking system, maintain the orphan source registration and retrieval program and upgrade the emergency preparedness