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)

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

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

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)

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

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)

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

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.)

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

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.

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

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

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

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

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)

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

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)

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)

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.)

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

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

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)

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

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

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

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.

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

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)

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

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)

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

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)

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)

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

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

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

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)

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)

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

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

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)

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

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

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.

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)

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

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

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

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

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

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)

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)

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)

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

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

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)

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)

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)

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

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.

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)

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

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.)

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)

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)

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)

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

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

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)

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

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)

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

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

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.

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)

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

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)

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

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

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

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)

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)

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)

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

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

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)

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)

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

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.

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

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)

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)

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

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)

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)

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)

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.

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

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

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%.

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)

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

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

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

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.

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

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)

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)

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

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)

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.

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)

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

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)

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)

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

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.

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

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

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)

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)

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)

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

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

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

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.)

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)

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

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.

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.

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

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

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

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

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

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.)

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

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)

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

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

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.

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

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

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

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.

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

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

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.

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

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.

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

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.

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)

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.

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)

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

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)

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

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

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)

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)

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)

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

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

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)

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

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)

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.)

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

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

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.)

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)

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

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)

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

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

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

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

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

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)

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

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)

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)

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

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

Probabilistic assessment of the Juragua NPP response under Station Blackout conditions

International Nuclear Information System (INIS)

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

1996-01-01

Assessment of the NPP response under SBO (station Blackout) conditions is a current safety issue of special interest, In the case of Juragua NPP, the safety assessment related to this topic is very important, taking into account the peculiarities of the Cuban Electro energetic System: small and long island, without possibilities of conexion beyond its borders and under the incidence of tropical phenomena In this papers a preliminary evaluation is presented of the potential incidence of Station Blackout conditions for Juragua NPP. the importance of this sort of events for the safety of the plant is evaluated, the factors which condition it are identified and measures for its prevention or recovering the normal situation if such an event takes place are proposed

Modernization of tritium in air monitoring system for CANDU type NPP

International Nuclear Information System (INIS)

Purghel, L.; Iancu, R.; Popescu, M.

2009-01-01

Nuclear energy provides at present one third of Europe's electricity with nearly no greenhouse-gas emissions. Sustained efforts are now being conducted to harmonize regulations all over Europe through WENRA and to converge on technical nuclear safety practices within the TSO network ETSON (European Technical Safety Organizations Network). In order to achieve this goals of safety function, IFIN-HH together with CITON developed a new solution to improve the Tritium Monitoring System (TMS) of Cernavoda NPP and the new generation of CANDU type reactors, using Tritium in air Intelligent Monitors (TIM) developed and patented by IFIN-HH. The paper presents a comparative analysis between the technical characteristics of traditional solutions implemented in CANDU type NPP, particularly in Cernavoda NPP Unit 2 and the newly proposed solution. (authors)

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)

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.)

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

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

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)

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)

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)

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)

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)

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)

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

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

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.)

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

Application of reliability centered maintenance to Embalse NPP

International Nuclear Information System (INIS)

Torres, Antonio; Perdomo, Manuel; Fornero, Damian; Corchera, Roberto

2010-01-01

One of the most recent applications of Probabilistic Safety Analysis to Embalse NPP is the Safety Oriented Maintenance Program developed through the Reliability Centered Maintenance (RCM) methodology. Such an application was carried out by a cooperated effort between the staff of nuclear safety department of NPP and experts from Instituto Superior de Tecnologias y Ciencias Aplicadas of Cuba. So far 6 technological systems have been analyzed with important results regarding the optimization of preventive and predictive maintenance program of those systems. Any tasks of RCM were automated via MOSEG code. The results of this study were focused on the elaboration and modification of the Preventive Program, prioritization of stocks, reorientation of predictive techniques and modification in the time parameters of maintenance. (author)

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

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)

Geological-Hydrological Site Evaluation for NPP Planning

Energy Technology Data Exchange (ETDEWEB)

Faust, Brigitte; Mini, Paolo [Nordostsschweizerische Kraftwerke AG NOK, Parkstrasse 23, 5401 Baden (Switzerland)

2008-07-01

NOK is investigating the potential replacement of the current NPP in Beznau. In order to meet the requirements with respect to a general licence application, geological, seismological, and geotechnical engineering, but also hydrological boundary conditions have been defined. These conditions define the nature of necessary investigations to obtain the geological, seismic, geotechnical and hydrological data themselves forming the basis to determine the site suitability. Viability has to be provided that a NPP can be built and operated at the proposed site without compromising public health, safety and environment. The collected data are also the basis for the design of all safety relevant structures, systems, and components. For example, the latter have to withstand the effects of natural phenomena such as earthquakes and human induced impact such as airplane crash without loosing their capability to perform the assigned safety functions. (authors)

Geological-Hydrological Site Evaluation for NPP Planning

International Nuclear Information System (INIS)

Faust, Brigitte; Mini, Paolo

2008-01-01

NOK is investigating the potential replacement of the current NPP in Beznau. In order to meet the requirements with respect to a general licence application, geological, seismological, and geotechnical engineering, but also hydrological boundary conditions have been defined. These conditions define the nature of necessary investigations to obtain the geological, seismic, geotechnical and hydrological data themselves forming the basis to determine the site suitability. Viability has to be provided that a NPP can be built and operated at the proposed site without compromising public health, safety and environment. The collected data are also the basis for the design of all safety relevant structures, systems, and components. For example, the latter have to withstand the effects of natural phenomena such as earthquakes and human induced impact such as airplane crash without loosing their capability to perform the assigned safety functions. (authors)

Analysis of an NPP Structure subjected to Vibrations Induced from Airplane Crashes

International Nuclear Information System (INIS)

Noh, Sang Hoon; Kim, Yong Soo; Kim, Chong Hak

2009-01-01

After the terrorists' attacks with civilian airplanes on September 11, 2001, special attention has been paid to the potential for an airplane crash into a Nuclear Power Plant (NPP) as a man-made hazard. An airplane crash (APC) into an NPP has the potential to damage the roofs and walls of these structures, as well as other systems and components such as pipelines, electric motors, power supplies, power cables of electricity transmission that are important for safety. Therefore, an evaluation of the structural response to an APC is important for the safety of NPPs to be confirmed. A structural integrity analysis was carried out focusing on the vibration effects of an APC on an NPP structure. The NPP structure under consideration has been conceptually redesigned based on APR1400 to have double containments for the purpose of a feasibility study to meet European requirements. The finite element method was used for the structural analysis of the NPP, and the computer code ABAQUS was employed for this analysis

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

Top level risk study for units 1 to 4 at NPP Kozloduy

International Nuclear Information System (INIS)

Shishkova, I.

1997-01-01

The first Top Level Study of NPP Kozloduy Units 1/4 was started in 1991 and completed in March 1992 by BEQE. The Primary objective of the project was to provide in a short term a tool for quantified comparison of the relative risk benefits that various design or operational modifications would have if implemented on NPP Kozloduy Units 1/4. This original version of the Study was reviewed by the IAEA who agreed in general with the conclusions indicating that the report provides valuable insights which should be used in the decision making process. The IAEA made suggestion for some changes. These comments were taken into account in the subsequent version. In early 1994 the new version was issued. The objective of this second project, in addition to revising the original study to reflect the IAEA comments, was to address the potential benefits from further improvements at NPP Kozloduy. The work on the second version was going on in parallel with the activities carried out in the mainframe of several NPP Kozloduy Units 1/4 Upgrading Program (WANO Six-Month Program, Program for the NPP Kozloduy Units 1-4 Upgrading, Twinning Program, etc.). And at last, a Top-Level Probabilistic Safety Study was carried out by EQE - Bulgaria and completed early in 1995 at the request of NPP Kozloduy. The objective of the project was the contribution of the different activities performed in the mainframe of the Program for the NPP Kozloduy Units 1-4 Safety Upgrading in the overall safety improvement to be assessed. 3 figs, 1 tab

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

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.

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

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)

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)

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

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

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)

Comparison of Domestic Safety Review and European Union(EU) Stress Test After Nuclear Accident in Fukushima Daiichi NPPs

Energy Technology Data Exchange (ETDEWEB)

Kang, Hwa Sung; Kim, Jin Weon [Chosun University, Gwangju (Korea, Republic of)

2016-05-15

The European Union(EU) nuclear regulators group established stress test criteria and procedures, and utilities performed a self-review in accordance with those criteria and procedures. For Wolsung nuclear unit-1,the stress test was additionally conducted for deciding the continued operation of NPP, even though the safety review had been conducted after Fukushima NPP accident. Thus, this study is to compares the process, criteria, and results of the safety review performed in domestic NPPs and EU stress test performed in Cernavoda NPP. From the comparisons, the effectiveness and necessity of the stress test to decide the continued operation of NPPs is discussed. and the improvement items for safety enhancement are derived. The comparison showed that the process and review criteria of EU stress test was more systematic and specific than those used in domestic NPPs. But it was indicated that the improvement items resulted from the safety review performed in domestic NPPs are more comprehensive and powerful than EU stress tests (Cernavoda NPP) results. EU stress test for Cernavoda NPP evaluated in 3 fieldsand derived 13 design change items. The 50 improvement items derived from domestic safety review were including the contents of these 13 items.

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

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)

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

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)

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)

Self-reliance and innovation of Qinshan phase II NPP project

International Nuclear Information System (INIS)

Ye Qizhen; Yang Lanhe

2007-01-01

This article mainly describes the self-reliance and innovation of Qinshan nuclear power project of phase II, in-between it contains new reactor core design, as well as related experimental and calculation analysis, especially for new reactor design produced fluid-induced vibration model test, theoretical analysis and testing in-built reactor; aiming at two-loop NSSS a series improvement made for safety systems and related safety analysis to enhance their reliability and redundancy; according to specialty of two-loop NSSS an optimization made for NPP parameters and design of related equipments, for the purpose to make the output of NPP maximal; design of main reactor building and T-G building also improved according to characteristics of two-loop NSSS and site conditions. CRDM and refueling machine are researched and manufactured on base of self-reliance, their performance are better than design requirements, large portion of key equipments are localized through different way. In construction first time realized the integrated erection of containment dome. During the commissioning non-nuclear steam driving of T-G set, as well as 500 kV high voltage rising using emergent diesel generator, etc. are carried out.In period of operation still continuous innovation and improvement are made, so that to keep the good record of operation. (authors)

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.

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)

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)

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.)

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)

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 '

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

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

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

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

NPP Information Model as an Innovative Approach to End-to-End Lifecycle Management of the NPP and Nuclear Knowledge Management Proven in Russia

International Nuclear Information System (INIS)

Tikhonovsky, V.; Kanischev, A.; Kononov, V.; Salnikov, N.; Shkarin, A.; Dorobin, D.

2016-01-01

Full text: Managing engineering data for an industrial facility, including integration and maintenance of all engineering and technical data, ensuring fast and convenient access to that information and its analysis, proves to be necessary in order to perform the following tasks: 1) to increase economic efficiency of the plant during its lifecycle, including the decommissioning stage; 2) to ensure strict adherence to industrial safety requirements, radiation safety requirements (in case of nuclear facilities) and environmental safety requirements during operation (including refurbishment and restoration projects) and decommissioning. While performing tasks 1) and 2), one faces a range of challenges: 1. A huge amount of information describing the plant configuration. 2. Complexity of engineering procedures, step-by-step commissioning and significant geographical distribution of industrial infrastructure. 3. High importance of plant refurbishment projects. 4. The need to ensure comprehensive knowledge transfer between different generations of operational personnel and, which is especially important for the nuclear energy industry, between the commissioning personnel generations. NPP information model is an innovative method of NPP knowledge management throughout the whole plant lifecycle. It is an integrated database with all NPP technical engineering information (design, construction, operation, diagnosing, maintenance, refurbishment). (author

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)

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

Nuclear Oversight Function at Krsko NPP

International Nuclear Information System (INIS)

Bozin, B.; Kavsek, D.

2010-01-01

The nuclear oversight function is used at the Krsko NPP constructively to strengthen safety and improve performance. Nuclear safety is kept under constant examination through a variety of monitoring techniques and activities, some of which provide an independent review. The nuclear oversight function at the Krsko NPP is accomplished by Quality and Nuclear Oversight Division (SKV). SKV has completed its mission through a combination of compliance, performance and effectiveness-based assessments. The performance-based assessment is an assessment using various techniques (observations, interviews, walk-downs, document reviews) to assure compliance with standards and regulations, obtain insight into performance, performance trends and also to identify opportunities to improve effectiveness of implementation. Generally, the performance-based approach to oversight function is based on some essential elements. The most important one which is developed and implemented is an oversight program (procedure). The program focuses on techniques, activities and objectives commensurate with their significance to plant operational safety. These techniques and activities are: self-assessments, assessments, audits, performance indicators, monitoring of corrective action program (CAP), industry independent reviews (such as IAEA's OSART and WANO Peer Review), industry benchmarking etc. Graded approach is an inherent product of a performance based program and ranking process. It is important not only to focus on the highest ranked performance based attributes but to lead to effective utilization of an oversight program. The attributes selected for oversight need to be based on plant specific experience, current industry operating experience, supplier's performance and quality issues. Collaboration within the industry and effective utility oversight of processes and design activities are essential for achieving good plant performance. So the oversight program must integrate relevant

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

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)

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)

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)

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

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...

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)

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

Preventing degradation and in-service inspection of NPP steam generator

International Nuclear Information System (INIS)

Ding Xunshen

1999-01-01

The author describes the degradation of steam generator tubes in initial operating stage of France NPP. The author emphatically presents the preventive measures to tackle degradation and In-service inspection of Daya Bay NPP 1 unit as a guarantee of safety operation, including secondary side water chemistry monitoring and controlling, leakage rate monitoring, eddy-current inspection, mechanical cleaning, cleanliness inspection, foreign objects removal and etc

Modernization and power upgrading of the Loviisa NPP

Energy Technology Data Exchange (ETDEWEB)

Keskinen, A. [IVO Power Engineering Ltd., Vantaa (Finland)

1997-12-31

In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.). 4 refs.

Modernization and power upgrading of the Loviisa NPP

International Nuclear Information System (INIS)

Keskinen, A.

1997-01-01

In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.)

Modernization and power upgrading of the Loviisa NPP

Energy Technology Data Exchange (ETDEWEB)

Keskinen, A [IVO Power Engineering Ltd., Vantaa (Finland)

1998-12-31

In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.). 4 refs.

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)

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

Fusion of eastern and western technology in VVER 1000 NPP upgrade

International Nuclear Information System (INIS)

Ubra, O.; Fleischhans, J.; Kveton, M.

1997-01-01

An extensive modernization program upgrading two units of VVER 1000 type of the Czech nuclear power plant (NPP) Temelin to meet the latest international standards is presented. The program is based primarily on combination of eastern and western technology and it has been implemented during plant construction. The NPP Temelin was originally designed according to the standards of the former Soviet Union. After a series of reviews in the 1990s, a decision was made by the Temelin management of upgrade the design of the plant, including the supply of fuel and instrumentation and control system by a western company. The adoption of western technology and practices has helped to solve a large number of IAEA safety issues related to design and operation of VVER 1000 NPP. Details on the current Temelin design and other related safety matters are presented

Atomic energy in Lithuania: Nuclear safety in 1997

International Nuclear Information System (INIS)

Kutas, S.; Krenevichius, R.; Chiuchelis, R.; Demchenko, M.

1998-01-01

Annual report of VATESI - Lithuanian Nuclear Safety Authority's activity in 1997 is presented. According to the prescribed responsibilities by the Law on Nuclear Energy and by the statute, VATESI's main fields of activities in 1997 were evaluation of the safety of Ignalina NPP, supervision of Ignalina NPP's operational safety, control of the implementation of safety improvement program SIP-2 in Ignalina NPP, preparation for the licensing of the unit No.1 of Ignalina NPP, accountancy and control of nuclear materials, regulation of radioactive waste management. Detailed description of all these activities is provided in the report. Important role of international assistance and cooperation is emphasized also. Year 1997 was safe for the operation of Ignalina NPP: there were no safety significant events according to the INES scale, only three events received qualification as level 1 events. In 1997 there was completed SAR of Ignalina NPP, its review report RSR and those were presented to the authorities. Taking into account all recommendations of SAR and RSR a new safety improvement program SIP-2 was made and started implemented

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)

The intelligent system for accident identification in NPP

International Nuclear Information System (INIS)

Hernandez, Jorge Luis.

1998-01-01

Accidental situations in NPP are of greet concern for operators, the facility, regulatory bodies and the environment. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making when 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 Analyses 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 system from the combination of the artificial intelligence techniques: fussy logic and artificial neural networks. The system works with variables from the process of the firsts 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

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)

Spent fuel pool risk analysis for the Dukovany NPP

Energy Technology Data Exchange (ETDEWEB)

Hust' ak, S.; Jaros, M.; Kubicek, J. [UJV Rez, a.s., Husinec-Rez (Czech Republic)

2013-07-01

UJV Rez, a.s. maintains a Living Probabilistic Safety Assessment (Living PSA) program for Dukovany Nuclear Power Plant (NPP) in the Czech Republic. This project has been established as a framework for activities related to risk assessment and to support for risk-informed decision making at this plant. The most extensively used PSA application at Dukovany NPP is risk monitoring of instantaneous (point-in-time) risk during plant operation, especially for the purpose of configuration risk management during plant scheduled outages to avoid risk significant configurations. The scope of PSA for Dukovany NPP includes also determination of a risk contribution from spent fuel pool (SFP) operation to provide recommendations for the prevention and mitigation of SFP accidents and to be applicable for configuration risk management. This paper describes the analysis of internal initiating events (IEs) in PSA for Dukovany NPP, which can contribute to the risk from SFP operation. The analysis of those IEs was done more thoroughly in the PSA for Dukovany NPP in order to be used in instantaneous risk monitoring. (orig.)

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.

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

A Study on Human Factors in Maintenance of a Nuclear Power Plant (NPP)

International Nuclear Information System (INIS)

Park, Young Ho; Seong, Poong Hyun

2006-01-01

In human factors research, more attention has been devoted to the operation of a nuclear power plant (NPP) than to their maintenance. However, more NPP incidents are caused by inadequate maintenance rather than by faulty operation. There is a trend in NPP toward introducing digital technology into safety and non-safety systems. This lead to changes of maintenance, and support systems such as diagnosis system, augmentation system and handy terminal will be developed. In this context, it is important to identify tasks of human related to each phase of maintenance and their relation in order to apply those to maintenance. However, there are few researches of human factors in maintenance. This paper studies on framework of cognitive task analysis for developing maintenance support systems

Atomic energy in Lithuania: Nuclear safety in 1996

International Nuclear Information System (INIS)

Vaishnys, P.; Krenevichius, R.; Alejev, A.; Demchenko, M.

1997-01-01

The first annual report of VATESI - Lithuanian Nuclear Safety Authority's activity is presented and cover description of the main results in 1996. VATESI was established in 1991 recently after regaining of independence when Ignalina NPP come into jurisdiction of Republic of Lithuania. Since the establishment in six year period of operation VATESI developed its activities considerably and now is able to perform the functions of independent regulator of nuclear safety in Lithuania. The main fields of VATESI activities in 1996 were evaluation of the safety of Ignalina NPP, supervision of Ignalina NPP's operational safety, control of the implementation of safety improvement program SIP-1 in Ignalina NPP, development of new regulatory legislation according to IAEA standards and practices, accountancy and control of nuclear materials, regulation of radioactive waste management. Detailed description of all these activities is provided in the report. In 1996 there were no safety significant events in Ignalina NPP according to the INES scale, only five events received qualification as level 1 events

Review of the main issues connected with the Belene NPP project

International Nuclear Information System (INIS)

Simeonov, P.

1996-01-01

The basic seismic studies connected with the construction of the second NPP in Bulgaria are outlined. The site selected for this plant is located on the right bank of the Donau River, near the Belene town. The seismicity of the site is investigated by the Bulgarian Academy of Sciences (BAS) and Energoproject Ltd., Sofia, and assessed by IAEA, Siemens, EQE and Westinghouse experts. The suggested coefficient for maximum seismic acceleration of 0.24 g takes into account the safety of the site, structures and equipment of the NPP. The conclusion made is that from seismic point of view the site is not hazardous. Construction works started in 1987 and were stopped in 1990 due to financial and socio-ecological reasons. The author's opinion is that the construction of the Belene NPP has no alternative and should be completed providing higher safety and reliability of the plant operation

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)

Monitoring of Persons at the Exit from Bohunice NPP

International Nuclear Information System (INIS)

Kaizer, J.; Svitek, J.

2001-01-01

Full text: IAEA defines the principal requirement 'defence in depth' as a multilayer system in its authorised document - International Basic Safety Standards for Protection against Ionising Radiation, Safety Series No. 115. The principle of the multilayer system is: a failure at one zone is compensated or corrected at subsequent zones. The main argument why Bohunice NPP modernised its monitoring system at the exit from NPP was the implementation of the principle 'defence in depth'. Several instruments PM7 (Eberline) equipped with the plastic scintillation detectors had been bought because of the modernisation. The instruments had to be integrated into overall security system NPP without any restriction to the number of passing people. The supplier had to modify the basic instrument operation. NPP required a 'dynamic' monitoring version, this means the operation without stopping of a person within the instrument. After the modification the value of the RDA (Reliable Detectable Activity) of the instruments PM7 was within the interval 9.25-10.4 kBq for 137 Cs (dotted source in the middle of the instrument). RDA for the mix of activation products was 2-3 times lower. In conclusion of our paper the results of the monitoring are presented within tree years as well as a discussion about these results. Maximum values of contaminations measured were very low (several kBq) that represented very low risk from potential exposure. (author)

External hazards analysis approach to level 1 PSA of Mochovce NPP - Slovakia

International Nuclear Information System (INIS)

Stojka, Tibor

2000-01-01

Analyses of external events had been first time performed at the design stage of the Mochovce NPP showing sufficiently low contribution of external hazards to core damage frequency. But, based on IAEA document 'Safety problems of WWER-440/213 NPPs and the categorization' (IAEA-EBP-WWER-03, 1996), the need of new reassessment arose due to discrepancy of some origin recommendations in compare with present IAEA ones. Mochovce NPP Nuclear Safety Improvements Program elaborated at the same time included the IAEA recommendations and following improvements were proposed to perform in context of external events. 1. Seismic project and new locality seismic evaluation This safety improvement includes also some 'on site' technical improvements in seismic stability of structures and equipment. 2. Unit specific analyses of extreme meteorologic conditions. This safety improvement focuses on impact of feasible extreme conditions on NPP systems caused by rain, snow and hail storms, frost, winds, low and high temperatures. 3. Analyses of external hazards caused by humans. In this safety improvement were specified: feasible sources of explosions; analyses of hydrogen, gas and propane-calor gas depots; air crash risk. The results of these implemented safety improvements were considered in the PSA study. The External hazards analysis is also part of Level 1 PSA Mochovce NPP performed by PSA Department of VUJE Trnava Inc., Engineering, Design and Research Organization, Slovakia. Some partial analyses are performed in cooperation with following companies DS and S - SAIC, USA and Geophysical Institute Academy of Science, Slovakia Relko, Slovakia. Basic documents are: NUREG/CR-2300 'PRA Procedures Guide - A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants' and IAEA SS No. 50-P-7 'Treatment of External Hazards in PSA for NPPs. The external hazards analysis consists of following parts: 1. Geography and plant locality; 2. Nearby industry; 3. Extreme

Hydrology and Oceanography Analysis Regarding The NPP Site Screening Process at Banten Province

International Nuclear Information System (INIS)

Yarianto-S-Budi-Susilo

2007-01-01

Regarding the NPP development in the future, it is needed to make inventory of potential site in the Java Island as well as in the outside Java Island. The NPP site inventory availability is to answer the energy demand challenge. Site screening process should be performed in accordance with the IAEA safety standard regarding the site selection, investigating several aspects related to the NPP safety (exclusion, safety and suitability factor) in the large area to obtain potential site candidates. For the site survey stage of hydrology and oceanography aspects, the analysis are more focused on the tidal phenomena along the north coastline, bathymetry, water resource, and hydrology system in the Banten Province. The method used are secondary data collection, field confirmation and internet searching. The result of the study showed that Tanjung Pujut and Tanjung Pasir are suitable based on the bathymetry and water intake facility consideration. Meanwhile Tanjung Kait and Tanjung Pasir more suitable considering tsunami aspects that may be generated by Krakatau Volcano. (author)

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

Innovative Approach to Implementation of FPGA-based NPP Instrumentation and Control Systems

Energy Technology Data Exchange (ETDEWEB)

Andrashov, Anton; Kharchenko, Vyacheslav; Sklyar, Volodymir [Centre for Safety Infrastructure-Oriented Research and Analysis, Kharkov (Ukraine); SIORA Alexander [Research and Production Corporation Radiy, Kirovograd (Ukraine)

2011-08-15

Advantages of application of Field Programmable Gates Arrays (FPGA) technology for implementation of Instrumentation and Control (I and C) systems for Nuclear Power Plants (NPP) are outlined. Specific features of FPGA technology in the context of cyber security threats for NPPs I and C systems are analyzed. Description of FPGA-based platform used for implementation of different safety I and C systems for NPPs is presented. Typical architecture of NPPs safety I and C system based on the platform, as well as approach to implementation of I and C systems using FPGA-based platform are discussed. Data on implementation experience of application of the platform for NPP safety I and C systems modernization projects are finalizing the paper.

Innovative approach to implementation of FPGA-based NPP instrumentation and control systems

International Nuclear Information System (INIS)

Andrashov, Anton; Kharchenko, Vyacheslav; Sklyar, Volodymir; Siora, Alexander

2011-01-01

Advantages of application of Field Programmable Gates Arrays (FPGA) technology for implementation of Instrumentation and Control (I and C) systems for Nuclear Power Plants (NPP) are outlined. Specific features of FPGA technology in the context of cyber security threats for NPPs I and C systems are analyzed. Description of FPGA-based platform used for implementation of different safety I and C systems for NPPs is presented. Typical architecture of NPPs safety I and C system based on the platform, as well as approach to implementation of I and C systems using FPGA-based platform are discussed. Data on implementation experience of application of the platform for NPP safety I and C systems modernization projects are finalizing the paper. (author)

Innovative Approach to Implementation of FPGA-based NPP Instrumentation and Control Systems

International Nuclear Information System (INIS)

Andrashov, Anton; Kharchenko, Vyacheslav; Sklyar, Volodymir; SIORA Alexander

2011-01-01

Advantages of application of Field Programmable Gates Arrays (FPGA) technology for implementation of Instrumentation and Control (I and C) systems for Nuclear Power Plants (NPP) are outlined. Specific features of FPGA technology in the context of cyber security threats for NPPs I and C systems are analyzed. Description of FPGA-based platform used for implementation of different safety I and C systems for NPPs is presented. Typical architecture of NPPs safety I and C system based on the platform, as well as approach to implementation of I and C systems using FPGA-based platform are discussed. Data on implementation experience of application of the platform for NPP safety I and C systems modernization projects are finalizing the paper

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

Kozloduy NPP - its 20 years operational experience; AEhS `Kozloduy`: 20-letni opyt ehksploatatsii

Energy Technology Data Exchange (ETDEWEB)

Yankov, M; Grizanov, I; Sabinov, S [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

1996-12-31

The main technical and economic results of 20 year operation of the Kozloduy NPP are summarised. The total electricity produced amounts to more than 200 billions KWh. The energy produced by the plant accounts for more than 40% of the national electricity output since 1993. Details of electricity production in different years by all 6 units are given. The efficiency of units 1-4 (WWER-440) is 29-30% and of units 5-6 (WWER-1000) is 29-33%. The Unit 6 has best characteristics. Since 1989 a safety test is being carried out in cooperation with the IAEA. Units 1 and 2 have been reconstructed in order to enhance safety and reliability. 1 tab.

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

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.)

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)

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

Nuclear safety. Improvement programme

International Nuclear Information System (INIS)

2000-01-01

In this brochure the improvement programme of nuclear safety of the Mochovce NPP is presented in detail. In 1996, a 'Mochovce NPP Nuclear Safety Improvement Programme' was developed in the frame of unit 1 and 2 completion project. The programme has been compiled as a continuous one, with the aim to reach the highest possible safety level at the time of commissioning and to establish good preconditions for permanent safety improvement in future. Such an approach is in compliance with the world's trends of safety improvement, life-time extension, modernisation and nuclear station power increase. The basic document for development of the 'Programme' is the one titled 'Safety Issues and their Ranking for WWER 440/213 NPP' developed by a group of IAEA experts. The following organisations were selected for solution of the safety measures: EUCOM (Consortium of FRAMATOME, France, and SIEMENS, Germany); SKODA Prague, a.s.; ENERGOPROJEKT Prague, a.s. (EGP); Russian organisations associated in ATOMENERGOEXPORT; VUJE Trnava, a.s

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

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)

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)

Advances in safety related maintenance

International Nuclear Information System (INIS)

2000-03-01

The maintenance of systems, structures and components in nuclear power plants (NPPs) plays an important role in assuring their safe and reliable operation. Worldwide, NPP maintenance managers are seeking to reduce overall maintenance costs while maintaining or improving the levels of safety and reliability. Thus, the issue of NPP maintenance is one of the most challenging aspects of nuclear power generation. There is a direct relation between safety and maintenance. While maintenance alone (apart from modifications) will not make a plant safer than its original design, deficient maintenance may result in either an increased number of transients and challenges to safety systems or reduced reliability and availability of safety systems. The confidence that NPP structures, systems and components will function as designed is ultimately based on programmes which monitor both their reliability and availability to perform their intended safety function. Because of this, approaches to monitor the effectiveness of maintenance are also necessary. An effective maintenance programme ensures that there is a balance between the improvement in component reliability to be achieved and the loss of component function due to maintenance downtime. This implies that the safety level of an NPP should not be adversely affected by maintenance performed during operation. The nuclear industry widely acknowledges the importance of maintenance in NPP safety and operation and therefore devotes great efforts to develop techniques, methods and tools to aid in maintenance planning, follow-up and optimization, and in assuring the effectiveness of maintenance

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)

New appraisement of siting for a NPP on Mures river

International Nuclear Information System (INIS)

Traian Mauna

2010-01-01

The studies for a second NPP siting on inner Romanian rivers began in a careful manner since 1982 as a first part of the Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. The experience gained from Cernavoda NPP siting, the first mission of new multi-branch of specialists team was to choose new NPP sites adapting the CANDU type NPP Cernavoda project to the new parameters of close water cooling circuit and of hard less or no rock foundation strata. The new sites conditions mean a lot of changes of CANDU license and a decrease the output power supplied to the national electric grid. The studies on the Mures river as alternative site of Olt river in Transylvania region began in 1986 and were stopped after 1990. This paper tries to reconsider shortly the old analysis focused on geological and geotechnical aspects and other local sites characteristics according to the last IAEA Safety Standards taking into account also the last types of NPP generations and the number of units. (author)

Pre feasibility assesment of smart off shore NPP (ONPP) of gravity based-structure type for Indonesia

International Nuclear Information System (INIS)

Sahala M Lumbanraja; Dharu Dewi

2017-01-01

The SMART ONPP GBS-type is a small power (100 MWe) pressurized water reactor, and located at offshore site. This technology was developed based on existing SMART nuclear technology & offshore drilling technology with a gravity-based type of structure. This is a response to the post-Fukushima accident, Japan (2011), to improve the safety system, overcome the land limitations, and minimize the public resistance to NPP cases in the inland. The purpose of this paper is to assess the pre-feasibility of the implementation of GBS NPP in Indonesia both in terms of technological feasibility and regulation. The method used is literature review and continued with descriptive analysis. The result shows that SMART ONPP are worth considering because they offer improved aspects of safety, offshore tread availability, and better public acceptance. So far, this NPP can not be implemented in Indonesia because it is hampered by Government Regulation No. 2 year 2014 regarding Licensing of Nuclear Installation Safety and Security which stipulates that site is an inland location and NPP built in Indonesia should be proven. (author)

Investigation of Chernobyl NPP and recovery of environment and medical service in Ukraine

International Nuclear Information System (INIS)

Narabayashi, Tadashi; Hamamoto, Kazuko; Mizumachi, Wataru; Okamoto, Koji

2013-01-01

Fukushima Daiichi NPP accident will be terminated, if sufficient care of recovery for heart of people, economy and environment leads to happy Fukushima. It might be taken more than 20 years or more. The overseas NPP investigation group in the study committee for safety regulatory of NPP in PES division in JSME visited Chernobyl NPP, Ministry of Emergencies, National Institute for Strategic Studies under the President of Ukraine, Chernobyl Center, Slavutych, Research Center for Radiation Medicine of the National Academy of Medical Sciences of Ukraine, National University of Life and Environmental Science of Ukraine, Institute for Safety Problems of Nuclear Power Plants and Chernobyl Museum. The investigation of bases of their new technology for deep LRW cleaning of Fukushima genesis from the whole range of radionuclides, including uranium and transuranic elements, organic impurities and simultaneous concentration of radioactive components in a small volume. The recovery of environment and care of heart of people were good in Ukraine. The lessons derived from the accident, we can decide what we should do. (author)

Activities of Nuclear Regulatory Authority and safety of nuclear facilities in the Slovak Republic in 1993

International Nuclear Information System (INIS)

1994-04-01

A brief account of activities carried out by the Nuclear Regulatory Authority of the Slovak Republic (NRA SR) in 1993 is presented. These activities are reported under the headings: (1) Introduction; (2) Regulatory activities at nuclear power plants units in operation; (2.1) Nuclear power plant SEP-EBO V-1; (4) Selected operation events and safety assessment in NPP SEP-EBO V-1; (2.2) Safety assessment of NPP SEP-EBO V-2; (3) Results of regulatory activities at the decommissioning of NPP A-1; (4) Regulatory activities at units under construction SEP-EMO - NPP Mochovce; (5) Further regulatory activities. (5.1) Preparation of designated personnel; (5.2) Inspection and accountancy of nuclear material; (5.3) Security provisions; (5.4) Accounted items and double use items; (5.5) Problem of radioactive wastes; (6.1) International co-operation activities of NRA; (6.2) Emergency planning; (6.3) International activities for quality enhancement of national supervision; (7) Conclusion [sk

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

Bohunice Nuclear Power Plant Safety Upgrading Program

International Nuclear Information System (INIS)

Toth, A.; Fagula, L.

1996-01-01

Bohunice nuclear Power Plant generation represents almost 50% of the Slovak republic electric power production. Due to such high level of commitment to nuclear power in the power generation system, a special attention is given to safe and reliable operation of NPPs. Safety upgrading and operational reliability improvement of Bohunice V-1 NPP was carried out by the Bohunice staff continuously since the plant commissioning. In the 1990 - 1993 period extensive projects were realised. As a result of 'Small Reconstruction of the Bohunice V-1 NPP', the standards of both the nuclear safety and operational reliability have been significantly improved. The implementation of another modifications that will take place gradually during extended refuelling outages and overhauls in the course of 1996 through 1999, is referred to as the Gradual Reconstruction of the Bohunice V-1 Plant. The general goal of the V-1 NPP safety upgrading is the achievement of internationally acceptable level of nuclear safety. Extensive and financially demanding modification process of Bohunice V-2 NPP is likely to be implemented after a completion of the Gradual Reconstruction of the Bohunice V-1 NPP, since the year 1999. With this in mind, a first draft of the strategy of the Bohunice V-2 NPP upgrading program based on Probabilistic Safety assessment consideration was developed. A number of actions with a general effect on Bohunice site safety is evident. All these activities are aimed at reaching the essential objective of Bohunice NPP Management - to ensure a safe, reliable and effective electric energy and heat generation at the Bohunice site. (author)

Automated personal dosimetry monitoring system for NPP

International Nuclear Information System (INIS)

Chanyshev, E.; Chechyotkin, N.; Kondratev, A.; Plyshevskaya, D.

2006-01-01

Full text: Radiation safety of personnel at nuclear power plants (NPP) is a priority aim. Degree of radiation exposure of personnel is defined by many factors: NPP design, operation of equipment, organizational management of radiation hazardous works and, certainly, safety culture of every employee. Automated Personal Dosimetry Monitoring System (A.P.D.M.S.) is applied at all nuclear power plants nowadays in Russia to eliminate the possibility of occupational radiation exposure beyond regulated level under different modes of NPP operation. A.P.D.M.S. provides individual radiation dose registration. In the paper the efforts of Design Bureau 'Promengineering' in construction of software and hardware complex of A.P.D.M.S. (S.H.W. A.P.D.M.S.) for NPP with PWR are presented. The developed complex is intended to automatize activities of radiation safety department when caring out individual dosimetry control. The complex covers all main processes concerning individual monitoring of external and internal radiation exposure as well as dose recording, management, and planning. S.H.W. A.P.D.M.S. is a multi-purpose system which software was designed on the modular approach. This approach presumes modification and extension of software using new components (modules) without changes in other components. Such structure makes the system flexible and allows modifying it in case of implementation a new radiation safety requirements and extending the scope of dosimetry monitoring. That gives the possibility to include with time new kinds of dosimetry control for Russian NPP in compliance with IAEA recommendations, for instance, control of the equivalent dose rate to the skin and the equivalent dose rate to the lens of the eye S.H.W. A.P.D.M.S. provides dosimetry control as follows: Current monitoring of external radiation exposure: - Gamma radiation dose measurement using radio-photoluminescent personal dosimeters. - Neutron radiation dose measurement using thermoluminescent

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)

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

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)

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)

Simulation of a turbine trip transient at Embalse NPP with full-circuit CATHENA model

Energy Technology Data Exchange (ETDEWEB)

Rabiti, A., E-mail: arabiti@na-sa.com.ar [Nucleoelectrica Argentina S.A., Embalse Nuclear Power Plant, Engineering Management Branch, Embalse (Argentina); Parrondo, A., E-mail: aparrondo@na-sa.com.ar [Nucleoelectrica Argentina S.A., Engineering Management, Buenos Aires (Argentina); Serrano, P., E-mail: pserrano@na-sa.com.ar [Nucleoelectrica Argentina S.A., Licensing Coordination Branch, Atucha II Project Branch (Unidad de Gestion), Buenos Aires (Argentina); Sablayrolles, A.; Damiani, H., E-mail: asablayrolles@na-sa.com.ar, E-mail: hdamiani@na-sa.com.ar [Nucleoelectrica Argentina S.A., Embalse Nuclear Power Plant, Embalse Life Extension Project Management, Embalse (Argentina)

2015-07-01

Embalse NPP is carrying on a Periodic Safety Review to deal with its life extension. This review includes tasks like Deterministic Analysis review for the Final Safety Analysis Report. In 2011, NA-SA (Nucleoelectrica Argentina S.A.) issued a first CATHENA full-circuit model representing the current plant. This model is used in this work. The simulation presented here corresponds to a turbine trip that occurred at Embalse NPP. Consistency between the simulation and the real event is demonstrated. Furthermore, NASA is currently performing Safety Analysis with a new model developed jointly with AECL and Candu Energy which includes post refurbishment changes and other improvements. (author)

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. �

Use of operator training and simulating complexes for development of NPP APCS

International Nuclear Information System (INIS)

Malkin, S.D.; Rakitin, I.D.; Sivokon', V.P.

2002-01-01

The evolution of NPP educational facilities - from systems of training of personnel to power simulating systems for the development of Automatic Process Control System (APCS) and safety analysis is followed. High efficiency of simulating systems application for the development of NPP APCS including optimization of man-machine system interface was shown by experiments of USA, France, Japan, United Kingdom, conforming examples were given [ru

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)

The dependence level analysis between the human actions in NPP Operation

International Nuclear Information System (INIS)

Farcasiu, M.; Nitoi, M.; Apostol, M.; Florescu, G.; Prisecaru, Ilie

2009-01-01

The Human Reliability Analysis (HRA) is an important method in Probabilistic Safety Assessment (PSA) studies and offers desirability for concrete improvement of the man - machine - organization interfaces, reliability and safety. An important step in HRA is the dependence level analysis between the human actions performed by the same person or between the actions performed by different persons, step in quantitative analysis of the human errors probabilities. The purpose of this paper is to develop a model to analyze the dependence level between human actions for Nuclear Power Plant (NPP) operation. The model estimates the conditional human error probabilities (CHEP) and joint human error probabilities (JHEP). The achieved sensitivity analyses determine human performance sensibility to systematic variations for dependence level between human actions. The human error probabilities estimated in this paper are adequate values for integration both in HRA and in PSA realized for NPP. This type of analysis helps in finding and analyzing the ways of reducing the likelihood of human errors, so that the impact of human factor to systems availability, reliability and safety can be realistically estimated. In order to demonstrate the usability of this model an analysis is performed upon the dependences between the necessary human actions in mitigating the consequences of LOCA events, particularly for the case of Cernavoda NPP. (authors)

Polish media and public opinion on NPP Mochovce commissioning

International Nuclear Information System (INIS)

Latek, Stanislaw

1999-01-01

The so called 'Mochovce Problem' was one of the major topics in Polish media in the period from May to July 1998. The nuclear power plant commissioning caused an unexpectedly strong reaction, especially in the newspapers, slightly less so in electronic media. Faced with clearly hostile media reaction to Mochovce NPP, the National Atomic Energy Agency representatives, together with atomic and nuclear experts, undertook to change these attitudes. In numerous interviews, letters to the editors and talks with journalists, they attempted to correct the mistakes, explain the true safety situation in the nuclear power plant, by whom it was constructed, who supervised and tested the systems and so on. The completion of Mochovce NPP construction improved significantly the electricity balance in Slovakia, thus decreasing the pressure for continuing the operation of older Bohunice V1 units beyond their design lifetime. For this reason, as well as in view of striving for improvement in environmental factors beyond Polish southern border, especially after Kyoto/97 decisions on greenhouse gases emissions, the public opinion in Poland should support the Mochovce NPP construction. In 1996 Poland has signed with Slovakia a bilateral inter-governmental agreement on the prompt notification on nuclear accidents and on the cooperation in the nuclear safety and radiological protection matters. On the basis of this agreement the experts from Polish National Atomic Energy Agency are in perpetual contact with Slovakian Nuclear Regulatory Body and in each and every moment can obtain full and comprehensive information on the plant parameters iportant for nuclear safety. The experts explanations, together with the NAEA top management visit to the plant itself, brought some results. The media became less aggressive, and Polish public and authorities - contrary to the Austrians - do not protest loudly against the commissioning of this newest European NPP. Now, in December 1998, the tune of

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

Operational safety performance indicators for nuclear power plants

International Nuclear Information System (INIS)

2000-05-01

Since the late 1980s, the IAEA has been actively sponsoring work in the area of indicators to monitor nuclear power plant (NPP) operational safety performance. The early activities were mainly focused on exchanging ideas and good practices in the development and use of these indicators at nuclear power plants. Since 1995 efforts have been directed towards the elaboration of a framework for the establishment of an operational safety performance indicator programme. The result of this work, compiled in this publication, is intended to assist NPPs in developing and implementing a monitoring programme, without overlooking the critical aspects related to operational safety performance. The framework proposed in this report was presented at two IAEA workshops on operational safety performance indicators held in Ljubljana, Slovenia, in September 1998 and at the Daya Bay NPP, Szenzhen, China, in December 1998. During these two workshops, the participants discussed and brainstormed on the indicator framework presented. These working sessions provided very useful insights and ideas which where used for the enhancement of the framework proposed. The IAEA is acknowledging the support and contribution of all the participants in these two activities. The programme development was enhanced by pilot plant studies. Four plants from different countries with different designs participated in this study with the objective of testing the applicability, usefulness and viability of this approach

Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

Energy Technology Data Exchange (ETDEWEB)

Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

2005-03-15

The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments.

Seismic response analysis of Wolsung NPP structure and equipment subjected to scenario earthquakes

International Nuclear Information System (INIS)

Choi, In Kil; Ahn, Seong Moon; Choun, Young Sun; Seo, Jeong Moon

2005-03-01

The standard response spectrum proposed by US NRC has been used as a design earthquake for the design of Korean nuclear power plant structures. However, it does not reflect the characteristic of seismological and geological of Korea. In this study, the seismic response analysis of Wolsung NPP structure and equipment were performed. Three types of input motions, artificial time histories that envelop the US NRC Regulatory Guide 1.60 spectrum and the probability based scenario earthquake spectra developed for the Korean NPP site and a typical near-fault earthquake recorded at thirty sites, were used as input motions. The acceleration, displacement and shear force responses of Wolsung containment structure due to the design earthquake were larger than those due to the other input earthquakes. But, considering displacement response increases abruptly as Wolsung NPP structure does nonlinear behavior, the reassessment of the seismic safety margin based on the displacement is necessary if the structure does nonlinear behavior; although it has adequate the seismic safety margin within elastic limit. Among the main safety-related devices, electrical cabinet and pump showed the large responses on the scenario earthquake which has the high frequency characteristic. This has great effects of the seismic capacity of the main devices installed inside of the building. This means that the design earthquake is not so conservative for the safety of the safety related nuclear power plant equipments

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)

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.)

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)

Main ways for solution of nuclear power plant safety in the USSR

International Nuclear Information System (INIS)

Sidorenko, V.A.; Kovalevich, O.M.; Kramerov, A.Ya.; Bagdasarov, Yu.E.

1977-01-01

In this paper the principles of provisions of nuclear power plant (NPP) safety in the USSR based on the accumulated scientific and engineering experience of design, construction and operating of NPP are discussed. Methods for the decision of the problem were different in various stages of development of nuclear energy and they will be changed in the light of future experience and when the scope of NPP utilization will be broadened. It was formulated the complex of means and ways for providing the real safety of NPP. It includes both the technical and organizing measures and all they are important to solve the NPP safety problem. The first two of these conditions of providing the safety - the high quality of NPP equipment and its constant control during the whole operation of NPP - have to minimize possible failures of power plant elements capable of resulting in an accident conditions. Design and utilization of effective safety systems at NPP and localization systems - these are the next two conditions to provide the safety called upon either to exclude or to attenuate up to the reasonable level the radioactive product yield outside the determined boundaries. The normalization of safety problems and the effective state inspection after keeping the required actions are the next necessary links of this complex called upon to ensure the fulfillment of all safety requirements. The report concerns the realization of formulated procedure to the provision of NPP safety with reactors of WWER and RBM-K types are being put into operation in the Soviet Union. The versatility of approach to the safety provision allows to attain the required safety by various means depending on the type of reactor, its siting and other features concentrating attention to one or another factors. Alongside with the above-mentioned, the paper considers the special requirements and safety measures regarding reactors of next generation such as liquid metal breeders of BN-type and other systems [ru

Safety performance indicators used by the Russian Safety Regulatory Authority in its practical activities on nuclear power plant safety regulation

International Nuclear Information System (INIS)

Khazanov, A.L.

2005-01-01

The Sixth Department of the Nuclear, Industrial and Environmental Regulatory Authority of Russia, Scientific and Engineering Centre for Nuclear and Radiation Safety process, analyse and use the information on nuclear power plants (NPPs) operational experience or NPPs safety improvement. Safety performance indicators (SPIs), derived from processing of information on operational violations and analysis of annual NPP Safety Reports, are used as tools to determination of trends towards changing of characteristics of operational safety, to assess the effectiveness of corrective measures, to monitor and evaluate the current operational safety level of NPPs, to regulate NPP safety. This report includes a list of the basic SPIs, those used by the Russian safety regulatory authority in regulatory activity. Some of them are absent in list of IAEA-TECDOC-1141 ('Operational safety performance indicators for nuclear power plants'). (author)

Role and responsibilities of management in NPP personnel training and competence. Working material

International Nuclear Information System (INIS)

1994-01-01

The main aim and result of this seminar was imparting knowledge to various levels of Paks NPP management on their special tasks and responsibilities to achieve personnel competence, which include: meeting relevant regulatory and other requirements; defining the qualifications for NPP personnel jobs; training using systematic approach to training to attain the required level of qualification and competence of all NPP personnel, which includes management, operations, maintenance and technical support personnel and others; recruiting and retaining qualified personnel, including career development; supporting the training of all personnel on their responsibilities for introducing, maintaining and improving safety. Refs, figs and tabs

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

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.

NPP long term operation in Spain - First application for license renewal

International Nuclear Information System (INIS)

Francia, L.; Gorrochategui, I.; Marcos, R.

2007-01-01

Full text: In the operation of the Spanish nuclear power plants (NPP), safety is always the prime consideration. Plant Life Management Programmes have been set up with the strategic objective to operate the NPPs as long as they are considered safe and reliable. The safety of each NPP is reviewed by the Spanish nuclear regulatory authority (CSN) under a continuous process. In addition, experience is gained from operating the plants and from exchanges with operators of similar units. Current Spanish regulatory framework for renewing NPP operating licenses requires performing a Periodic Safety Review (PSR) to be performed every 10 years and submitted when applying for a new renewal of the NPP operating license. A few years ago, CSN issued a document regarding the licensing requirements that nuclear power plants should meet in order to be granted with an operating license for long term operation (i.e, operation beyond the original plant design life, typically 40 years). Besides the traditional PSR requirements, specific requirements regarding to long term operation (LTO) include: - An Aging Management and Evaluation Program, including the identification and evaluation of Time Limited Aging Analysis (TLAA). - An updated Radiological Impact Study. - A review and assessment of regulation/standard applicability. Garona NPP (GE, BWR/3 design) operated by Spanish utility Nuclenor from 1971 has a current operating license up to 2009. A decision was made to apply for a new operating license, being Garona plant the first one in Spain to face with the new long term operation requirements. The paper will provide an overview of the methodology used in Spain to address and perform the required analyses to support the LTO application for the operating license renewal. In particular, focus will be paid on the project developed in Garona (2002-2006) whose result has been the first Spanish application for License Renewal for LTO. Also it will be reported the ongoing work necessary to

Radiation monitoring in the NPP environment, control of radioactivity in NPP-environment system

International Nuclear Information System (INIS)

Egorov, Yu.A.

1987-01-01

Problems of radiation monitoring and control of the NPP-environment system (NPPES) are considered. Radiation control system at the NPP and in the environment provides for the control of the NPP, considered as the source of radioactive releases in the environment and for the environmental radiation climate control. It is shown, that the radiation control of the NPP-environment system must be based on the ecological normalization principles of the NPP environmental impacts. Ecological normalization should be individual for the NPP region of each ecosystem. The necessity to organize and conduct radiation ecological monitoring in the NPP regions is pointed out. Radiation ecological monitoring will provide for both environmental current radiation control and information for mathematical models, used in the NPPES radiation control

Improved Nuclear Power Plant maintenance. Application to Cofrentes NPP

International Nuclear Information System (INIS)

Serradell, V.

2002-01-01

The Industrial Safety and Environment Group (MEDASEGI) of the Valencia Polytechnic University (UPV) develops its activities in the fields of environmental radioactivity metrology and risk analysis and management aimed at improving industrial safety. The different lines of research developed by this group include collaboration with Cofrentes NPP. This collaboration began in 1993 and has primarily focused on issues related to improved maintenance and currently on life management. (Author)

Technology for NPP decantate treatment realized at Kola NPP

International Nuclear Information System (INIS)

Stakhiv, Michael; Avezniyazov, Slava; Savkin, Alexander; Fedorov, Denis; Dmitriev, Sergei; Kornev, Vladimir

2007-01-01

At Moscow SIA 'Radon' jointly with JSC 'Alliance Gamma', the technology for NPP Decantate Treatment was developed, tested and realized at Kola NPP. This technology consists of dissolving the salt residue and subsequent treatment by ozonization, separation of the deposits formed from ozonization and selective cleaning by ferro-cyanide sorbents. The nonactive salt solution goes to an industrial waste disposal site or a repository specially developed at NPP sites for 'exempt waste' products by IAEA classification. This technology was realized at Kola NPP in December 2006 year. At this time more than 1000 m 3 of decantates log time stored are treated. It allows solving very old problem to empty decantates' tanks at NPPs in environmentally safe manner and with high volume reduction factor. (authors)

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)

Problem of dynamic impacts upon NPP buildings and constructions

International Nuclear Information System (INIS)

Kosterev, A.E.

1998-01-01

Activities of IAEA member-states dealing with the problems of taking into account external natural and technogenic impacts (earthquakes, cyclones, hurricanes, aircraft crushes, air shock waves) when designing NPP buildings and constructions important for nuclear and radiation safety are briefly discussed. It is concluded that development of IAEA general recommendations for taking into account specific dynamic impacts will improve noticeably safety of NPPs being in operation both in Russia and all over the World as a result of reconstruction of safety-important buildings and constructions

Relationship towards Engineering, Quality Assurance and 10CFR50.59 in the Design Change Process at the Krsko NPP

International Nuclear Information System (INIS)

Basic, I.; Krajnc, B.

1998-01-01

The paper discusses the relationship between the Krsko NPP design change engineering practice, quality assurance and the USA Nuclear Regulatory Commission 10 Code Federal Rule 50.59 (10CFR50.59). Together these controls ensure that plant design bases are maintained and yield a safe design. The 10CFR50.59 has been applied in Krsko NPP plant specific procedure entitled ESP-2.303 ''Authorization of Changes, Tests and Experiments'' (Safety Evaluation Screening) since 1994. All proposed changes requiring Safety Evaluations are being submitted to the SNSA (Slovenian Nuclear Safety Administration). If the proposed change is constituting an ''Unreviewed Safety Question'' the formal licensing procedure shall be completed before design change can be implemented otherwise the proposed design change is rejected. The procedure(ESP-2.303) provides the methodology to be followed in determining if a proposed activity involves an unreviewed safety question. An ''Unreviewed Safety Question'' is essentially the same as defined in 10CFR50.59(a)(2): ''A proposed change, test or experiment shall be deemed to involve an unreviewed safety question (1) if the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the (updated) safety analysis report may be increased; or (2) if a possibility for an accident or malfunction of a different type than any evaluated previously in the (updated) safety analysis report may be created; or (3) if the margin of safety as defined in the basis for any technical specification is reduced.'' This paper discusses the Following Krsko NPP Safety Evaluation aspects: 1. Defense in Depth Design Philosophy 2. Methodology 3. Definitions and Applicability of Terms 4. Evaluation Process Guidance and Documentation Process 5. Krsko NPP Lessons Learned. (author)

Seismic site evaluation practice and seismic design guide for NPP in Continent of China

Energy Technology Data Exchange (ETDEWEB)

Yuxian, Hu [State Seismological Bureau, Beijing, BJ (China). Inst. of Geophysics

1997-03-01

Energy resources, seismicity, NPP and related regulations of the Continent of China are briefly introduced in the beginning and two codes related to the seismic design of NPP, one on siting and another on design, are discussed in some detail. The one on siting is an official code of the State Seismological Bureau, which specifies the seismic safety evaluation requirements of various kinds of structures, from the most critic and important structures such as NPP to ordinary buildings, and including also engineering works in big cities. The one on seismic design of NPP is a draft subjected to publication now, which will be an official national code. The first one is somewhat unique but the second one is quite similar to those in the world. (author)

Seismic site evaluation practice and seismic design guide for NPP in Continent of China

International Nuclear Information System (INIS)

Hu Yuxian

1997-01-01

Energy resources, seismicity, NPP and related regulations of the Continent of China are briefly introduced in the beginning and two codes related to the seismic design of NPP, one on siting and another on design, are discussed in some detail. The one on siting is an official code of the State Seismological Bureau, which specifies the seismic safety evaluation requirements of various kinds of structures, from the most critic and important structures such as NPP to ordinary buildings, and including also engineering works in big cities. The one on seismic design of NPP is a draft subjected to publication now, which will be an official national code. The first one is somewhat unique but the second one is quite similar to those in the world. (author)

Increase nuclear safety of WWER-440

International Nuclear Information System (INIS)

Nochev, T.; Sabinov, S.

2000-01-01

A complete program for increasing nuclear safety has been made at NPP Kozloduy with the participation of German, French, Russian and American specialists. This effort cost greater than 100 mil $. This report includes the methods of increasing nuclear safety. The style of management in NPP Kozloduy has been changed for the last seven years. (authors)

Planning, conduct and principal features of NPP emergency exercises in Switzerland

International Nuclear Information System (INIS)

Baggenstos, M.

1993-01-01

Emergency exercises for each NPP are required on a regular basis by the Swiss Nuclear Safety Inspectorate. The purpose of such exercises is to train the NPP staff and the on-site emergency organization in the application of the emergency procedures and the cooperation with off-site emergency teams and public authorities. The paper discusses the purpose of the emergency exercises and experiences made especially with bilateral exercises. The responsibilities for the preparation and execution of the different emergency exercises in Switzerland are explained

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

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)

Units 3 and 4 of Kozloduy NPP - modernized in compliance with the contemporary international standards

International Nuclear Information System (INIS)

Urutchev, V.

2005-01-01

In this paper the Kozloduy NPP units 1- 4 original design characteristics are given. The main results from units 3 and 4 modernization programme as well as: assurance of reactor core safety; assurance of SLA reliability; assurance of unit safety and reliability are presented. The Kozloduy NPP units 3 and 4 modernization programme and the position of the units in the generally accepted scale of risk assessment are illustrated. The fulfillment of the IAEA and EC requirements and AQG recommendations is also presented. At the end the author concluded that: 1) The results from units 3 and 4 large scale modernization programs realization are assessed by international community as one of the most effective examples of good international practices applying in the filed of the nuclear facilities safety; 2) There are not any technical reasons behind discussions for early closure of units 3 and 4; 3) Taking into account the Balkans electricity consumption prognosis and the condition of the electricity production facilities, Kozloduy NPP units 3 and 4 can have sufficient contribution for the region reliable electricity delivery in the future years

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

Commissioning of Temelin NPP as seen by scientific supervisory group

International Nuclear Information System (INIS)

Svoboda, C.

2003-01-01

Scientific Supervisory Group worked during the Temelin NPP commissioning process as an independent supervisor. The main tasks and main results of its activity are described in this contribution. The characteristic common features of commissioning process and most important events from the Scientific Supervisory Group point of view are presented. In April 1999 the Czech Power Utility with the objective to achieve maximum level of nuclear safety and quality within the NPP Temelin commissioning procedures has established a special body / Scientific Supervisory Group and requested Nuclear Research Institute Rez plc to perform the required function. The Scientific Supervisory Group proceeds in accordance with its Statute and provides an independent specialised professional and expert work focused on nuclear safety assurance, assesment of the selected documentation related to plant preparedness for the individual commissioning stages, and, of course. on assessment of the commissioning tests results. While performing its function the Scientific Supervisory Group is guided by the Atomic Act and the relevant Directives of State Office for Nuclear Safety; its activities are in compliance with the applicable IAEA recommendations (Authors)

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)

NPP Evaluation, backfitting and life extension. An engineering viewpoint

Energy Technology Data Exchange (ETDEWEB)

Gonzalez Lopez, A [Empresarios Agrupados, A.I.E., Madrid (Spain)

1993-12-15

During the decade of the 80s, the Owners of the two oldest operating plants in Spain designed and built during the 60s - namely, Jose Cabrera NPP, a Westinghouse PWR, and Santa Maria de Garona NPP, a GE BWR- undertook the following important programs: 1. A far-reaching Systematic Evaluation Program (SEP) for the Jose Cabrera NPP consisting in the systematic safety review of the plant design, followed by the necessary hardware modifications, to upgrade it and make it comply with current safety criteria, and a Plant Upgrading Program for the Garona Nuclear Station focusing on specific topics affecting GE BWR Mark-I type plants of the same vintage. 2. A Remaining Life Management Program to ensure that the units, after extensive backfittings and high capital investment, would complete their design life, leaving open the option for plant life extension. These two units are today considered by the Spanish nuclear industry as the pilot plants for Plant Life Extension (PLEX) programs for PWRs and BWRs in our country The purpose of this paper is to summarize the principal lessons learned from EMPRESARIOS AGRUPADOS' participation as an architect-engineering organization in the engineering, design and implementation of these Programs. They are practical examples of positive experience which could be considered as a reference when carrying out similar programs for other plants. (author)

NPP Evaluation, backfitting and life extension. An engineering viewpoint

International Nuclear Information System (INIS)

Gonzalez Lopez, A.

1993-01-01

During the decade of the 80s, the Owners of the two oldest operating plants in Spain designed and built during the 60s - namely, Jose Cabrera NPP, a Westinghouse PWR, and Santa Maria de Garona NPP, a GE BWR- undertook the following important programs: 1. A far-reaching Systematic Evaluation Program (SEP) for the Jose Cabrera NPP consisting in the systematic safety review of the plant design, followed by the necessary hardware modifications, to upgrade it and make it comply with current safety criteria, and a Plant Upgrading Program for the Garona Nuclear Station focusing on specific topics affecting GE BWR Mark-I type plants of the same vintage. 2. A Remaining Life Management Program to ensure that the units, after extensive backfittings and high capital investment, would complete their design life, leaving open the option for plant life extension. These two units are today considered by the Spanish nuclear industry as the pilot plants for Plant Life Extension (PLEX) programs for PWRs and BWRs in our country The purpose of this paper is to summarize the principal lessons learned from EMPRESARIOS AGRUPADOS' participation as an architect-engineering organization in the engineering, design and implementation of these Programs. They are practical examples of positive experience which could be considered as a reference when carrying out similar programs for other plants. (author)

Results and development prospects for the Cernavoda NPP

International Nuclear Information System (INIS)

Rotaru, Ioan

2001-01-01

The new competitive market in the power sector offers favorable conditions for continual developing of Nuclear Power in Romania. The principal arguments in this respect are: the CANDU technology is a western technology recognized world wide both for its operation reliability and its economical performances. The results obtained by Cernavoda NPP Unit 1 over a period of four years and a half commercial operation confirms the validity of the CANDU project in Romania in the new conditions of competitive market. Accomplishing the Unit 2 of Cernavoda NPP represents for the time being one of the most cost-effective investment in the Romanian power sector. Also, the participation of the local industry to the project implementation is important as it ensures the initial heavy water and nuclear fuel charging, specific materials and equipment production, design and construction works, etc. Heaving in view all these favorable arguments, the construction of Unit 2 of Cernavoda NPP represents one of the highest priorities of the State Budget of the present administration in year 2001. The Nuclear Power Project in Romania, based on CANDU technology meets national as well as EU regulations concerning nuclear power and as such its development poses no significant problems both in the context of EU extension and in the frame of the developing competitive market in the power sector. For SNN SA the main objectives in view for the next stage are: operating Cernavoda NPP Unit 1 in conditions of highest nuclear safety and economical efficiency; finalizing the Cernavoda NPP Unit 2 and creation of an organisation for operating the two units; promoting of a functional system characteristic to the competitive market economy and open to privatization; creating opportunities for exporting electrical energy. The contribution contains the following sections: 1. General overlook; 2. Performance of the power production in the year 2000; 2.1. Unit 1 of Cernavoda NPP; 2.2. Nuclear Fuel Plant

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

Key features of MIR.1200 (AES-2006) design and current stage of Leningrad NPP-2 construction

International Nuclear Information System (INIS)

Ivkov, Igor

2010-01-01

MIR.1200/AES-2006 is an abbreviated name of the evolving NPP design developed on the basis of the VVER-1000 Russian design with gross operation life of 480 reactor-years. This design is being implemented in four Units of Leningrad NPP-2 (LNPP-2. The AES-91/99 was used as reference during development of the AES-2006 design for LNPP-2; this design was implemented in two Units of Tianwan NPP (China). The main technical features of the MIR.1200/AES-2006 design include a double containment, four trains of active safety systems (4x100%, 4x50%), and special engineering measures for BDBA management (core catcher, H2 PARs, PHRS) based mainly on passive principles. The containment is described in detail, the main features in comparison with the reference NPP are outlined, the design layout principles are highlighted, the safety system structure and parameters are described. Attention is paid to the BDBA management system, hydrogen removal system, core catcher, and PHRS-SG and C-PHRS. (P.A.)

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.

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.)

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.

The national experience of using the concept of the safety culture and its possible adaptation in the environmental field

International Nuclear Information System (INIS)

Barbashev, S.V.; Skalets'kij, Yu.M.; Voronenko, V.V.

2011-01-01

Based on interdisciplinary research a successful attempt of conceptual combination of ecological culture and ecological security is made. Because of the concept of the safety culture, an attempt to eliminate further restrictions on enhance NPP safety related to a person is made. The main feature of the safety culture is in the fact that the level of duties should be higher and better than usual good practice.

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.

Nuclear safety activities in the SR of Slovenia in 1986

Energy Technology Data Exchange (ETDEWEB)

Susnik, J [Inst. Jozef Stefan, Ljubljana (Slovenia)

1987-06-15

Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1986. (author)

Nuclear safety activities in the SR of Slovenia in 1986

International Nuclear Information System (INIS)

Susnik, J.

1987-06-01

Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1986. (author)

Ignalina Safety Analysis Group's report for the year 1998

International Nuclear Information System (INIS)

Uspuras, E.; Augutis, J.; Bubelis, E.; Cesna, B.; Kaliatka, A.

1999-02-01

Results of Ignalina NPP Safety Analysis Group's research are presented. The main fields of group's activities in 1998 were following: safety analysis of reactor's cooling system, safety analysis of accident localization system, investigation of the problem graphite - fuel channel, reactor core modelling, assistance to the regulatory body VATESI in drafting regulations and reviewing safety reports presented by Ignalina NPP during the process of licensing of unit 1

Follow-up of the 1993 ASSET mission to Borssele NPP

International Nuclear Information System (INIS)

Wieman, J.L.

1996-01-01

A description is given of the proposed in-house application of ASSET guidelines for periodic assessments of the Borssele NPP safety performance, and the definition of management initiatives to improve this performance. The periodic assessments will provide a continuation of work performed by the IAEA ASSET mission of 1993. The presentation is intended for experience exchange only, and does not report an authorized or formalized safety review. Figs, 3 tabs

Follow-up of the 1993 ASSET mission to Borssele NPP

Energy Technology Data Exchange (ETDEWEB)

Wieman, J L [NV, EPZ, Vlissingen (Netherlands)

1997-12-31

A description is given of the proposed in-house application of ASSET guidelines for periodic assessments of the Borssele NPP safety performance, and the definition of management initiatives to improve this performance. The periodic assessments will provide a continuation of work performed by the IAEA ASSET mission of 1993. The presentation is intended for experience exchange only, and does not report an authorized or formalized safety review. Figs, 3 tabs.

Safety upgrading at PAKS Nuclear Power Plant

International Nuclear Information System (INIS)

Bajsz, J.; Elter, J.

2000-01-01

The operation of Paks NPP has reached its half time. Until this time the plant fulfilled expectations raised before its construction: the four units have produced safely and reliably more than 200 TWh electricity. The production of the plant has been at the stable level since its construction and has provided 43-38 % of electricity consumed in Hungary. The annual production is around 14 TWh, which means a load factor higher than 85 %. Safety upgrading activities [1] at Paks had started in the late eighties, when the commissioning work of units 3 and 4 were carried out. That time the main emphases were put to lessons learned of the TMI and Chernobyl accidents. The international reviews hosted by our plant widened our review's scope. To systematize our approach a complete safety review, the AGNES (Advanced General Safety and New Evaluation of Safety) project was started in 1991. The goal of the project was to evaluate to what extent Paks NPP satisfied the current international safety expectations and to help in determining the priorities for safety enhancement and upgrading measures. The project completed in 1994 ranked our safety upgrading measures by safety significance, which became a basis for technical design work and financial scheduling. The other important outcome of the AGNES project was the introduction the Periodical Safety Review regime by our nuclear authority. These periodical reviews held after 10 years of operation offer the possibility - and obligation for the licensee - to perform a comprehensive assessment of the safety of the plant, to evaluate the integral effects of changes of circumstances happened during the review period. The goal of these reviews is to deal with cumulative effects of NPP ageing, modifications, operating experience and technical developments aimed at ensuring a high level of safety throughout plant service life. The execution of our safety-upgrading program is well advancing. For the whole program from 1996 to 2002 250

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)

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

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

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

An Analysis of Cyber-Attack on NPP Considering Physical Impact

Energy Technology Data Exchange (ETDEWEB)

Lee, In Hyo; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Son, Han Seong [Joonbu University, Geumsan (Korea, Republic of)

2016-05-15

Some research teams performed related works on cyber-physical system which is a system that cyber-attack can lead to serious consequences including product loss, damage, injury and death when it is attacked. They investigated the physical impact on cyber-physical system due to the cyber-attack. But it is hard to find the research about NPP cyber security considering the physical impact or safety. In this paper, to investigate the relationship between physical impact and cyber-attack, level 1 PSA results are utilized in chapter 2 and cyber-attack analysis is performed in chapter 3. The cyber security issue on NPP is inevitable issue. Unlike general cyber security, cyber-physical system like NPP can induce serious consequences such as core damage by cyber-attack. So in this paper, to find how hacker can attack the NPP, (1) PSA results were utilized to find the relationship between physical system and cyber-attack and (2) vulnerabilities on digital control systems were investigated to find how hacker can implement the possible attack. It is expected that these steps are utilized when establishing penetration test plans or cyber security drill plans.

An Analysis of Cyber-Attack on NPP Considering Physical Impact

International Nuclear Information System (INIS)

Lee, In Hyo; Kang, Hyun Gook; Son, Han Seong

2016-01-01

Some research teams performed related works on cyber-physical system which is a system that cyber-attack can lead to serious consequences including product loss, damage, injury and death when it is attacked. They investigated the physical impact on cyber-physical system due to the cyber-attack. But it is hard to find the research about NPP cyber security considering the physical impact or safety. In this paper, to investigate the relationship between physical impact and cyber-attack, level 1 PSA results are utilized in chapter 2 and cyber-attack analysis is performed in chapter 3. The cyber security issue on NPP is inevitable issue. Unlike general cyber security, cyber-physical system like NPP can induce serious consequences such as core damage by cyber-attack. So in this paper, to find how hacker can attack the NPP, (1) PSA results were utilized to find the relationship between physical system and cyber-attack and (2) vulnerabilities on digital control systems were investigated to find how hacker can implement the possible attack. It is expected that these steps are utilized when establishing penetration test plans or cyber security drill plans

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)

Prioritic directions of inculcation of diagnostic equipment at NPP

International Nuclear Information System (INIS)

Morozov, V.I.

2000-01-01

The diagnostic provision creates the conditions for increasing the safety and reliability of the NPP functioning, technical service and maintenance by the actual state. With an account of the large number of the NPP equipment elements, limitedness of financial resources, different technical-economical effect from diagnostics determination of the priority directions for introduction of technical diagnostic means into the operational practice is one of the main factors. The method for determining the above-mentioned priorities is proposed. The main aspects of the method and mathematical models, based on the logical-probabilistic modeling, are presented. The essence of the method consists in ranging the technical-economical effect from introduction of various factors of the diagnostic equipment [ru

Core Management of NPPs in China and Experience of Tianwan NPP

International Nuclear Information System (INIS)

Guohan, G.

2006-01-01

The Current Status of the NPPs in China; the Nuclear Power Development Plan till 2020; the Fuel Management and the Core Management Strategy of NPPs are briefly presented in this paper. The gained experience during construction of Taiwan NPP, new design features and major construction milestones of Taiwan NPP are also discussed. The full list of 21 non-conformance reports (NCR) submitted by Jiangsu Nuclear Power Company to NNSA during construction of Taiwan NPP is given. The main issues raised during FSAR review like: RP operation with incomplete number of primary loops; Sedimentation in the construction and piping of service water system; Location of postulated pipe break; Stratification analysis for surge line of pressurizer; Safety classification of equipment in primary circuit; Fire resistant capability of cable insulation; Fire Hazard Analysis; Optimization of protection against ionizing radiation ect. are also listed

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.

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

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

A way of Dukovany NPP to privatization and liberalized market

International Nuclear Information System (INIS)

Kouklik, I.

2001-01-01

This presentation describes the current situation in the Dukovany NPP with two main upcoming future phenomena - privatization of CEZ company and preparation for liberalized electricity market. Considerations about level of safety, investment costs and competitiveness of Dukovany Nuclear Power Plant in liberalization process is described together with a comparison of some safety features of the world's NPPs. Results of this comparison are used for consideration and evaluation of some require modifications effectiveness. (author)

The reevaluation of seismic safety of existing nuclear power plant

International Nuclear Information System (INIS)

Kitagawa, Hiroshi; Tominaga, Shohei; Kumagai, Chiyoshi; Koshiba, Koremutsu; Kono, Tomonori; Agawa, Kazuyoshi; Kuwata, Kenichiro

2003-01-01

We have carried out additional geological surveys in order to enrich our database on geological faults in the vicinity of Shimane Nuclear Power Plant (NPP). Prior to additional geological surveys, given the social importance of nuclear power plants, we hypothetically assumed that almost the whole length of an area covered by surveys would be an active fault that must be considered in seismic design, and tried to reevaluate the seismic safety of the NPP by applying an input earthquake ground motion larger than the level at the design stage. As a result, we have confirmed that seismic safety of the NPP can be maintained. This paper describes the method that we employed to reevaluate the seismic safety of Shimane NPP. (author)

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)

Environmental qualification program for Krsko NPP

International Nuclear Information System (INIS)

Cerjak, J.; Klenovsek, P.; Pavsek, J.; Freeland, K.R.; Spalj, S.

1998-01-01

NEK plant components, including those critical to safe plant operation, deteriorate and wear over service life due to the effects of aging and harsh environmental conditions. Since the plant environment is a source of common-cause failures, an Environmental Qualification (EQ) program is required to ensure and demonstrate the ability of safety-related equipment to perform its design safety function during a design-basis (DBE), even after aging over its service life in the plant. EQ is a requirement for plants licensed by the US NRC, in accordance with 10 CFR 50.49, Regulatory Guide 1.89, NUREG-0588, and IEEE-323. This paper presents the current EQ Program status at Krsko NPP.(author)

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

Integrated risk assessment for multi-unit NPP sites—A comparison

International Nuclear Information System (INIS)

Kumar, C. Senthil; Hassija, Varun; Velusamy, K.; Balasubramaniyan, V.

2015-01-01

Highlights: • Framework for integrated risk assessment for multi-unit NPP sites. • Categorization of external and internal events. • Modelling of key issues: mission time, cliff-edge, common cause failures, etc. • Safety goals for multi-unit NPP sites. • Comparison of site core damage frequency in one, two, three and four unit sites. - Abstract: Most of the nuclear power producing sites in the world houses multiple units. Such sites are faced with hazards generated from external events: earthquake, tsunami, flood, etc. and can threaten the safety of nuclear power plants. Further, risk from a multiple unit site and its impact on the public and environment was evident during the Fukushima nuclear disaster in March 2011. It is therefore important to evolve a methodology to systematically assess the risk from multi-unit site. For a single unit site, probabilistic risk assessment technique identifies the potential accident scenarios, their consequences, and estimates the core damage frequency that arise due to internal and external hazards. This challenging task becomes even more complex for a multiple unit site, especially when the external hazards that has the potential to generate one or more correlated hazards or a combination of non-correlated hazards are to be modelled. This paper presents an approach to evaluate risk for multiple NPP sites and also compare the risk for sites housing single, double and multiple nuclear plants.

Integrated risk assessment for multi-unit NPP sites—A comparison

Energy Technology Data Exchange (ETDEWEB)

Kumar, C. Senthil, E-mail: cskumar@igcar.gov.in [AERB-Safety Research Institute, Kalpakkam (India); Hassija, Varun; Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Balasubramaniyan, V. [AERB-Safety Research Institute, Kalpakkam (India)

2015-11-15

Highlights: • Framework for integrated risk assessment for multi-unit NPP sites. • Categorization of external and internal events. • Modelling of key issues: mission time, cliff-edge, common cause failures, etc. • Safety goals for multi-unit NPP sites. • Comparison of site core damage frequency in one, two, three and four unit sites. - Abstract: Most of the nuclear power producing sites in the world houses multiple units. Such sites are faced with hazards generated from external events: earthquake, tsunami, flood, etc. and can threaten the safety of nuclear power plants. Further, risk from a multiple unit site and its impact on the public and environment was evident during the Fukushima nuclear disaster in March 2011. It is therefore important to evolve a methodology to systematically assess the risk from multi-unit site. For a single unit site, probabilistic risk assessment technique identifies the potential accident scenarios, their consequences, and estimates the core damage frequency that arise due to internal and external hazards. This challenging task becomes even more complex for a multiple unit site, especially when the external hazards that has the potential to generate one or more correlated hazards or a combination of non-correlated hazards are to be modelled. This paper presents an approach to evaluate risk for multiple NPP sites and also compare the risk for sites housing single, double and multiple nuclear plants.

An intelligent and integrated V and V environment design for NPP I and C software systems

International Nuclear Information System (INIS)

Koo, Seo Ryong; Son Han Seong; Seong, Poong Hyun

2001-01-01

Nuclear Power Plant (NPP) is the safety critical system. Since, nuclear instrumentation and control (I and C) systems including the plant protection system play the brain part of human, nuclear I and C systems have an influence on safety and operation of NPP. Essentially, software V and V should be performed for the safety critical systems based on software. It is very important in the technical aspect because of the problems concerning license acquisitions. In this work, an intelligent and integrated V and V environment supporting the automation of V and V was designed. The intelligent and integrated V and V environment consists of the intelligent controller part, components part, interface part, and GUI part. These parts were integrated systematically, while taking their own independent functions

A Computer Program for Assessing Nuclear Safety Culture Impact

Energy Technology Data Exchange (ETDEWEB)

Han, Kiyoon; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of)

2014-10-15

Through several accidents of NPP including the Fukushima Daiichi in 2011 and Chernobyl accidents in 1986, a lack of safety culture was pointed out as one of the root cause of these accidents. Due to its latent influences on safety performance, safety culture has become an important issue in safety researches. Most of the researches describe how to evaluate the state of the safety culture of the organization. However, they did not include a possibility that the accident occurs due to the lack of safety culture. Because of that, a methodology for evaluating the impact of the safety culture on NPP's safety is required. In this study, the methodology for assessing safety culture impact is suggested and a computer program is developed for its application. SCII model which is the new methodology for assessing safety culture impact quantitatively by using PSA model. The computer program is developed for its application. This program visualizes the SCIs and the SCIIs. It might contribute to comparing the level of the safety culture among NPPs as well as improving the management safety of NPP.

Current status of safety analysis report for ANPP

International Nuclear Information System (INIS)

Amirjanyan, A.

1999-01-01

Current situation concerning Armenian NPP safety analysis report is considered within the frame of accepted safety practice. Licensing procedure is being developed. Technical support group was established in the Armenian Nuclear Regulatory Authority (ANRA). The task of the group is to study modern methods of NPP in depth safety analysis for technical assistance for the ANRA, and perform independent safety assessments. ANRA will be obliged to demand assistance from various foreign organisations for preparation of different parts of the Safety Analysis Report like determination though certain parts can be prepared in Armenia

Experimental analysis of specification language impact on NPP software diversity

International Nuclear Information System (INIS)

Yoo, Chang Sik; Seong, Poong Hyun

1998-01-01

When redundancy and diversity is applied in NPP digital computer system, diversification of system software may be a critical point for the entire system dependability. As the means of enhancing software diversity, specification language diversity is suggested in this study. We set up a simple hypothesis for the specification language impact on common errors, and an experiment based on NPP protection system application was performed. Experiment result showed that this hypothesis could be justified and specification language diversity is effective in overcoming software common mode failure problem

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)

Report on nuclear safety on the operation of nuclear facilities in 1989

International Nuclear Information System (INIS)

Gregoric, M.; Levstek, M. F.; Horvat, D.; Kocuvan, M.; Cresnar, N.

1990-01-01

Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1989.

Report on nuclear safety on the operation of nuclear facilities in 1990

International Nuclear Information System (INIS)

Gregoric, M.; Grlicarev, I.; Horvat, D.; Levstek, M.F.; Lukacs, E.; Kocuvan, M.; Skraban, A.

1991-06-01

Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1990.

Cernavoda NPP Unit 1 - a plant of several generations

International Nuclear Information System (INIS)

Rotaru, I.; Metes, M.; Anghelescu, M.S.

2001-01-01

The paper reflects some key aspects related to the shift of generations during the project's development, including the present stage. Further, the place of Cernavoda NPP Unit 1 in the Romanian power sector and among other nuclear stations in the world is presented. The operational performances achieved 'in service' up to the end of 1999, with reference to the performance indicators for electrical energy production, nuclear safety, radiation protection, radioactive wastes and nuclear fuel are illustrated. For all of these items, comparisons are performed with similar indicators reported by other worldwide nuclear power plants, in order to assess our results. Finally, some comments about Cernavoda NPP Unit 2 project status and need to completion and commissioning it are included. (authors)

Integrated assessment of radio-ecological situation using Gis-technology and spatial data bank of environment of NPP 30-km area

International Nuclear Information System (INIS)

Lev, T.D.; Tishchenko, O.G.; Piskun, V.N.

2011-01-01

In accordance with the requirements of IAEA on providing information-analytical and cartographic materials for NPP emergency systems Institute for safety problems of NPP is creating the spatial data bank taking into account geographical features, socio-economic indexes and information of monitoring of NPP environment. Based on analysis of connection between ecological parameters and geographical features of locality the specific of radio-ecological conditions of territories of NPP 30-m area is exposed by the example of Rivne and Khmel'nickiy NPP. Determination on critical territories maximally influencing on forming of the radiation dose for a population will allow to develop in good time preventive and protective measures for a population at the different scenarios of accident situation.

Multimedia system for the visitors' centre at the Ignalina NPP

International Nuclear Information System (INIS)

Alvers, Margareta

1999-01-01

The contents illustrated with video clips, animations, photographs, show the follwing: History of Ignalina NPP (INPP) growing; Visaginas - how the town came into being; Lake Druksiai; Development of nuclear power; Technical data of INPP; Description of INPP; Characteristic features of RBMK reactors; Reactor design; Technical parametres of RBMK-1500 reactor; Nuclear reaction and nuclear fission; Types of nuclear reactors; Circuits and systems; Radiation safety; Safety systems at the INPP; Upgrading nuclear safety at INPP following the Chernobyl accident; Safety problems at MP; Radioactive waste management in the world; RW Management at MP; Energy in Lithuania (thermal power stations, cogeneration plants, producing biogas from organic waste)

Development of Information Datasheets of Nuclear Power Plant (NPP) Equipment using cfiXLM schema

International Nuclear Information System (INIS)

Lee, Jaiho; Song, Eunhye

2014-01-01

In 2009, EPRI (Electrical Power Research Institute) published a new NPP information handover guide to provide NPP owners and operators with data handover templates in consistent format for effective delivery of information during all stages of the handover process. Another difficult concern for NPP data information management is to exchange the data information among many organizations such as NPP owners, operators, engineering companies, suppliers, and vendors. As a matter of fact, the improperly formatted handover of information sometimes occurs due to the discrepancy of data format (e. g., data description language type). This improper delivery can make negative effects on NPP integrity and safety. Thus, the lack of proper exchange for different data information systems of organizations should be resolved by using an international standard data format. The standard data format can reduce the cost and time for data exchange in each phase for design, procurement, delivery, installation, operation and maintenance of equipment. The AEX(automating equipment information exchange) pilot implementation project team under EPRI advanced nuclear technology (ANT) program has been conducted a research for the use of XML equipment schemas for electronic data exchange(EDE). They applied XML equipment schema for the design, selection, quotation, purchase and mock install of a safety injection centrifugal pump using EDE standard HI(hydraulic institute) 50.7. For data exchange, FIATECH, an industry consortium, has equally developed library of templates and reference data for ISO-15926, which is an international standard capable of reducing data-error and delivery time for exchanging data among different organizations. KHNP as an only owner/operator company has not experienced much difficulty in data interoperability with other organizations, but continued its unremitting exertions to develop a robust system capable of managing data information generated in all the stages of NPP

Development of Information Datasheets of Nuclear Power Plant (NPP) Equipment using cfiXLM schema

Energy Technology Data Exchange (ETDEWEB)

Lee, Jaiho; Song, Eunhye [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

2014-05-15

In 2009, EPRI (Electrical Power Research Institute) published a new NPP information handover guide to provide NPP owners and operators with data handover templates in consistent format for effective delivery of information during all stages of the handover process. Another difficult concern for NPP data information management is to exchange the data information among many organizations such as NPP owners, operators, engineering companies, suppliers, and vendors. As a matter of fact, the improperly formatted handover of information sometimes occurs due to the discrepancy of data format (e. g., data description language type). This improper delivery can make negative effects on NPP integrity and safety. Thus, the lack of proper exchange for different data information systems of organizations should be resolved by using an international standard data format. The standard data format can reduce the cost and time for data exchange in each phase for design, procurement, delivery, installation, operation and maintenance of equipment. The AEX(automating equipment information exchange) pilot implementation project team under EPRI advanced nuclear technology (ANT) program has been conducted a research for the use of XML equipment schemas for electronic data exchange(EDE). They applied XML equipment schema for the design, selection, quotation, purchase and mock install of a safety injection centrifugal pump using EDE standard HI(hydraulic institute) 50.7. For data exchange, FIATECH, an industry consortium, has equally developed library of templates and reference data for ISO-15926, which is an international standard capable of reducing data-error and delivery time for exchanging data among different organizations. KHNP as an only owner/operator company has not experienced much difficulty in data interoperability with other organizations, but continued its unremitting exertions to develop a robust system capable of managing data information generated in all the stages of NPP

TSO assistance towards the improvement of nuclear safety in Lithuania: achievements and perspectives

International Nuclear Information System (INIS)

Butcher, P.; Bystedt, P.; Chouha; Weber, J.P.; Zilys, K.

2001-01-01

This paper explores the improvements that have taken place at INPP (Ignalina nuclear power plant) in both the engineering and operational aspects of nuclear safety, discussing as example some areas where there has been a large involvement of international experts. The development of the Lithuanian Nuclear Regulatory Authority VATESI, and also the technical support organisations for both the regulator and the NPP are discussed. In each of these sections the paper describes some of the successes and also the areas where there have been some problems. Many of the problems have been resolved but challenges remain for the future. The paper only deals with the assistance towards enhancing the safety of the NPP up to the time of its closure. Regulation of decommissioning is an important area for the future and is receiving urgent attention and active planning at the present time. (authors)

Human factor in the problem of Russian nuclear industry safety

International Nuclear Information System (INIS)

Abramova, V.

2002-01-01

The approach to human factor definition, considered in the paper, consists of recognition of as many as possible factors for developing a complete list of factors, which have influence on mistakes or successful work of NPP personnel. Safety culture is considered as the main factor. The enhancement in nuclear power industry includes an optimization of organizational structures and development of personnel safety attitudes. The organizational factors, as possible root causes for human errors, need to be identified, assessed and improved. The organizational activities taken in Russia are presented

Changing NPP consumption patterns in the Holocene: from Megafauna "liberated" NPP to "ecological bankruptcy"

Science.gov (United States)

Doughty, C.

2015-12-01

There have been vast changes in how net primary production (NPP) is consumed by humans and animals during the Holocene beginning with a potential increase in availability following the Pleistocene megafauna extinctions. This was followed by the development of agriculture which began to gradually restrict availability of NPP for wild animals. Finally, humans entered the industrial era using non-plant based energies to power societies. Here I ask the following questions about these three energy transitions: 1. How much NPP energy may have become available following the megafauna extinctions? 2. When did humans, through agriculture and domestic animals, consume more NPP than wild mammals in each country? 3. When did humans and wild mammals use more energy than was available in total NPP in each country? To answer this last question I calculate NPP consumed by wild animals, crops, livestock, and energy use (all converted to units of MJ) and compare this with the total potential NPP (also in MJ) for each country. We develop the term "ecological bankruptcy" to refer to the level of consumption where not all energy needs can be met by the country's NPP. Currently, 82 countries and a net population of 5.4 billion are in the state of ecologically bankruptcy, crossing this threshold at various times over the past 40 years. By contrast, only 52 countries with a net population of 1.2 billion remain ecologically solvent. Overall, the Holocene has seen remarkable changes in consumption patterns of NPP, passing through three distinct phases. Humans began in a world where there was 1.6-4.1% unclaimed NPP to consume. From 1700-1850, humans began to consume more than wild animals (globally averaged). At present, >82% of people live in countries where not even all available plant matter could satisfy our energy demands.

Probabilistic safety assessment for instrumentation and control systems in nuclear power plants. A literature survey

International Nuclear Information System (INIS)

Lu, Lixuan; Jiang, Jin

2003-01-01

Deregulation in electricity market will create a great deal of challenges for Nuclear Power Plants (NPP). To stay competitive, NPP will need to find new ways to reduce their operation costs. In NPP, Instrumentation and Control (I and C) systems play an important role in reducing the cost of producing electricity while maintaining and/or enhancing safety. Therefore, it is extremely important that one should manage the I and C systems more efficiently and economically. Meanwhile, obsolescence problem associated with I and C systems encouraged the usage of advanced digital techniques in I and C systems. Thus, new methodologies are needed to analyze the reliability and determine the maintenance strategy for the digital I and C systems. Probabilistic Safety Assessment (PSA) has been probed to be a promising method to deal with this issue. This paper provides a literature survey on the development of digital I and C systems in NPP, followed by a detailed review of PSA including its benefits, limitations and the future direction of its development. Most importantly, potential applications of PSA in various aspects of I and C systems are brought into perspective throughout the paper. Furthermore, the applicability of PSA in the regulation of safety-related I and C systems is demonstrated. Detailed information on PSA applications in 1) the resource allocation for I and C systems: 2) the determination of surveillance testing strategies; and 3) I and C system designs, is provided. (author)

Application of the Concept of Intrusion Tolerant System for Evaluating Cyber Security Enhancements

Energy Technology Data Exchange (ETDEWEB)

Lee, Chanyoung; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

2016-10-15

One of the major problems is that nuclear industry is in very early stage in dealing with cyber security issues. It is because that cyber security has received less attention compared to other safety problems. In addition, late adoption of digital I and C systems has resulted in lower level of cyber security advancements in nuclear industry than ones in other industries. For the cyber security of NPP I and C systems, many regulatory documents, guides and standards were already published. These documents include cyber security plans, methods for cyber security assessments and comprehensive set of security controls. However, methods which can help assess how much security is improved if a specific security control is applied are not included in these documents. Hence, NPP I and C system designers may encounter difficulties when trying to apply security controls with limited structure and cost. In order to provide useful information about cyber security issues including cyber security enhancements, this paper suggests a framework to evaluate how much cyber security is improved when a specific cyber security enhancement is applied in NPPs. In order to provide useful information about cyber security issues including cyber security enhancements, this paper suggests a framework to evaluate how much cyber security is improved when a specific cyber security enhancement is applied in NPPs. The extent of cyber security improvement caused by security enhancement was defined as reduction ratio of the failure probability to secure the system from cyber-attack as Eq.1. The concept of 'intrusion tolerant system' was applied to not only prevent cyber-attacks but also limit the extent of damage in this study. For applying the concept of intrusion tolerant system to NPP, the event tree was constructed with some assumptions. Cyber security improvement caused by cyber security enhancement can be estimated as Eq.3. By comparing current system to the enhanced system, it is

Application of the Concept of Intrusion Tolerant System for Evaluating Cyber Security Enhancements

International Nuclear Information System (INIS)

Lee, Chanyoung; Seong, Poong Hyun

2016-01-01

One of the major problems is that nuclear industry is in very early stage in dealing with cyber security issues. It is because that cyber security has received less attention compared to other safety problems. In addition, late adoption of digital I and C systems has resulted in lower level of cyber security advancements in nuclear industry than ones in other industries. For the cyber security of NPP I and C systems, many regulatory documents, guides and standards were already published. These documents include cyber security plans, methods for cyber security assessments and comprehensive set of security controls. However, methods which can help assess how much security is improved if a specific security control is applied are not included in these documents. Hence, NPP I and C system designers may encounter difficulties when trying to apply security controls with limited structure and cost. In order to provide useful information about cyber security issues including cyber security enhancements, this paper suggests a framework to evaluate how much cyber security is improved when a specific cyber security enhancement is applied in NPPs. In order to provide useful information about cyber security issues including cyber security enhancements, this paper suggests a framework to evaluate how much cyber security is improved when a specific cyber security enhancement is applied in NPPs. The extent of cyber security improvement caused by security enhancement was defined as reduction ratio of the failure probability to secure the system from cyber-attack as Eq.1. The concept of 'intrusion tolerant system' was applied to not only prevent cyber-attacks but also limit the extent of damage in this study. For applying the concept of intrusion tolerant system to NPP, the event tree was constructed with some assumptions. Cyber security improvement caused by cyber security enhancement can be estimated as Eq.3. By comparing current system to the enhanced system, it is possible to

Test research and analysis for ultimate capacity of Qinshan NPP PCCV

International Nuclear Information System (INIS)

Zufeng, X.

1994-01-01

This paper introduces design and research for containment of Qinshan NPP which is the first PWR in CHINA designed and constructed by ourselves. The PCCV design is basically in conformity to ASME code. To verify the structural integrity capacity of Qinshan NPP containment, we fulfilled SIT and ILRT successfully in June, 1991. The special attention of the paper is focused on the ultimate capacity of the PCCV under severe accidents and earthquake. A study comprised of five different independent parts has been performed for the development of containment model test and corresponding nonlinear analysis. There are two prestressed concrete containment models with equipment hatch. One is 1/15 scale with steel liner tested on shake table and then moved out loaded with atmospheric pressure. The other is 1/10 scale without steel liner loaded with water pressure until destruction. From different methods including model test and nonlinear analysis, all obtained unanimous conclusion. The capacity under internal pressure and earthquake is reliable. The safety margin is enough. Consequently, in the second phase of Qinshan NPP and other PWR NPP under design, PCCV should be a better selection in China since it's more economic, rational and safe. (author)

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)

Assessment of Human Performance and Safety Culture at the Paks Nuclear Power Plant

International Nuclear Information System (INIS)

Toth, Janos; Hadnagy, Lajos

2002-01-01

Evaluation of human performance and safety culture of the personnel at a Nuclear Power Plant is a very important element of the self assessment process. At the Paks NPP a systematic approach to this problem started in the early 90's. The first comprehensive analysis of the human performance of the personnel was performed by the Hungarian Research Institute for Electric Power (VEIKI). The analysis of human failures is also a part of the investigation and analysis of safety related reported events. This human performance analysis of events is carried out by the Laboratory of Psychology of the plant and a supporting organisation namely the Department of Ergonomics and Psychology of the Budapest University of Technical and Economical Sciences. The analysis of safety culture at the Paks NPP has been in the focus of attention since the implementation of the INSAG-4 document started world-wide. In 1993 an IAEA model project namely 'Strengthening Training for Operational Safety' was initiated with a sub-project called 'Enhancement of Safety Culture'. Within this project the first step was the initial assessment of the safety culture level at the Paks NPP. It was followed by some corrective actions and safety culture improvement programme. In 1999 the second assessment was performed in order to evaluate the progress as a result of the improvement programme. A few indicators reflecting the elements of safety culture were defined and compared. The assessment of the safety culture with a survey among the managers was performed in September 2000 and the results are being evaluated at the moment. The intention of the plant management is to repeat the assessment every 2-3 years and evaluate the trend of the indicator. (authors)

Analysis of transients for NPP with VVER-440 using the code SiTAP

International Nuclear Information System (INIS)

Kalinenko, V.

1994-06-01

The report contains analysis of transients ''Loop connection'' and ''Steam generator tube rupture'' for nuclear power plants (NPP) with VVER-440. To obtain more detailed information about NPP's dynamic characteristics, various variants of initial and boundary conditions are considerd. Calculation of these transients was performed using the SiTAP code developed at the Nuclear Safety Institute of the Russian Research Centre ''Kurchatov Institute''. SiTAP code is a multifunctional computer tool for fast analysis of transient and accidental processes of VVER type reactors for engineers working in the field of NPP dynamics. SiTAP can be used form comparative analysis of several variants of accident scenarios to find out the conditions leading to most serious consequences from a safety point of view. In such cases, additional analyses using best-estimate codes should be carried out. The results of SiTAP for a faulty loop connection leading to a boron dilution accident are intended to be used as boundary conditions for a more detailed anlaysis with the aid of the three-dimensional reactor core model DYN3D, developed in the Research Centre Rossendorf for the simulation of reactivity initiated accidents. (orig.)

PHWR safety: design, siting and construction

International Nuclear Information System (INIS)

Sharma, V.K.

2002-01-01

In all activities associated with NPPs viz. siting, design, construction, commissioning and operation, safety is given overriding importance. The safety design principles of PHWRs are based on defence-in-depth approach, physical and functional separation between process and safety systems and also among various safety systems, redundancy to meet single failure criteria and postulation of a number of design basis events for which the plant must be designed. Apart from engineered safety systems, PHWRs have inherent characteristics which contribute to safety. In siting of a NPP, it is required to ensure that the given site does not pose undue radiological hazard to public and the environment both during normal operation as well as during and following an accident condition. For this purpose, all site related external events, both natural and man induced, are assessed for their effect on the plant and are considered as part of the design basis. Possible radiological impact of the NPP on environment and surrounding population is assessed and ensured to be within acceptable limits. During construction phase, it is essential that the NPP be built in accordance with design intent and with required quality of workmanship to ensure that the NPP will remain safe during all states of operation. This is achieved through careful execution and QA activities encompassing all aspects of component fabrication at manufacturer works, civil construction, site erection, assembly, and commissioning. Future trends in nuclear safety will continue to be based on existing principles which have proved to be sound. These will be further strengthened by features such as increasing use of passive means of performing safety functions and a more explicit treatment of severe accidents. (author)

Chapter No.4. Safety analyses

International Nuclear Information System (INIS)

2002-01-01

In 2001 the activity in the field of safety analyses was focused on verification of the safety analyses reports for NPP V-2 Bohunice and NPP Mochovce concerning the new profiled fuel and probabilistic safety assessment study for NPP Mochovce. The calculation safety analyses were performed and expert reviews for the internal UJD needs were elaborated. An important part of work was performed also in solving of scientific and technical tasks appointed within bilateral projects of co-operation between UJD and its international partnership organisations as well as within international projects ordered and financed by the European Commission. All these activities served as an independent support for UJD in its deterministic and probabilistic safety assessment of nuclear installations. A special attention was paid to a review of probabilistic safety assessment study of level 1 for NPP Mochovce. The probabilistic safety analysis of NPP related to the full power operation was elaborated in the study and a contribution of the technical and operational improvements to the risk decreasing was quantified. A core damage frequency of the reactor was calculated and the dominant initiating events and accident sequences with the major contribution to the risk were determined. The target of the review was to determine the acceptance of the sources of input information, assumptions, models, data, analyses and obtained results, so that the probabilistic model could give a real picture of the NPP. The review of the study was performed in co-operation of UJD with the IAEA (IPSART mission) as well as with other external organisations, which were not involved in the elaboration of the reviewed document and probabilistic model of NPP. The review was made in accordance with the IAEA guidelines and methodical documents of UJD and US NRC. In the field of calculation safety analyses the UJD activity was focused on the analysis of an operational event, analyses of the selected accident scenarios

Industry Operating Experience Process at Krsko NPP

International Nuclear Information System (INIS)

Bach, B.; Bozin, B.; Cizmek, R.

2012-01-01

Experience has shown that number of minor events and near misses, usually without immediate or significant impact to plant safety and reliability, are precursors of significant or severe events due to the same or similar root or apparent cause(s). It is therefore desirable to identify and analyze weaknesses of the precursor problems (events) in order to prevent occurrence of significant events. Theoretically, significant events could be prevented from occurring if the root cause(s) of these precursor problems could be identified and eliminated. The Operating Experience Program identifies such event precursors and by reporting them to the industry, plant specific corrective actions can be taken to prevent events at other operational plants. The intent of the Operating Experience Program is therefore to improve nuclear power plant safety and reliability of the operating nuclear power plants. Each plant develops its own Operating Experience Program in order to learn from the in-house operating experience as well as from the world community of nuclear plants. The effective use of operating experience includes analyzing both plant and industry events in order to identify fundamental weaknesses and then determining appropriate plant-specific actions that will minimize the likelihood of similar events. Learning and applying the lessons from operating experience is an integral part of station safety culture and is encouraged by managers throughout the top plant administrative programs and procedures. Krsko NPP is developed it own Operating Experience Program by using the most relevant INPO/WANO/IAEA guidelines as well as its own knowledge, skills an operating practice. The Operating Experience Program is a part of the Corrective Action Program, which is among top management programs, thus program is strongly encouraged by top management. The purpose of Operating Experience Program is to provide guidance for using, sharing, and evaluating operating experience information

NPOESS Preparatory Project (NPP) Science Overview

Science.gov (United States)

Butler, James J.

2011-01-01

NPP Instruments are: (1) well understood thanks to instrument comprehensive test, characterization and calibration programs. (2) Government team ready for October 25 launch followed by instrument activation and Intensive Calibration/Validation (ICV). NPP Data Products preliminary work includes: (1) JPSS Center for Satellite Applications and Research (STAR) team ready to support NPP ICV and operational data products. (2) NASA NPP science team ready to support NPP ICV and EOS data continuity.

Numerical Methods for an Analysis of Hydrogen Behaviors Coupled with Thermal Hydraulics in a NPP Containment

International Nuclear Information System (INIS)

Kim, Jongtae; Park, Rae-Joon; Hong, Seong-Wan; Kim, Gun-Hong

2016-01-01

In a containment safety analysis, multi-dimensional characteristics in thermal hydraulics are very important because the flow paths are not confined in a large free volume of the containment. The analysis is difficult because of a difference in length scales between a characteristic length of the flow and representative length of the containment. In order to simulate hydrogen and steam behaviors in a containment during postulated severe accidents, the GASFLOW code as a multi-dimensional analysis tool for NPP containment has been used for years because of its computational efficiency. Though GASFLOW is well developed for a real NPP containment analysis, there exist shortcomings in nodalization, two-phase and turbulence models. It is based on a Cartesian or cylindrical coordinate mesh, so it is impractical to refine a mesh locally in a region with a physical or geometrical complication. In this paper, the importance of the hydrogen safety in an NPP containment and requirements of the analysis tool was described. And physical models necessary for the hydrogen safety analysis code were listed. As a member of international collaborative project HYMERES for containment thermal hydraulics, KAERI is actively participating in an analytic working group. As an analysis tool for blind benchmarkes, the analysis code described in this paper was used. From the blind benchmark analyses, it was found that the code is very promising for hydrogen safety analysis. Currently, it is proposed to develop the code collaboratively in a hydrogen safety community based on an open-source strategy

Numerical Methods for an Analysis of Hydrogen Behaviors Coupled with Thermal Hydraulics in a NPP Containment

Energy Technology Data Exchange (ETDEWEB)

Kim, Jongtae; Park, Rae-Joon; Hong, Seong-Wan; Kim, Gun-Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In a containment safety analysis, multi-dimensional characteristics in thermal hydraulics are very important because the flow paths are not confined in a large free volume of the containment. The analysis is difficult because of a difference in length scales between a characteristic length of the flow and representative length of the containment. In order to simulate hydrogen and steam behaviors in a containment during postulated severe accidents, the GASFLOW code as a multi-dimensional analysis tool for NPP containment has been used for years because of its computational efficiency. Though GASFLOW is well developed for a real NPP containment analysis, there exist shortcomings in nodalization, two-phase and turbulence models. It is based on a Cartesian or cylindrical coordinate mesh, so it is impractical to refine a mesh locally in a region with a physical or geometrical complication. In this paper, the importance of the hydrogen safety in an NPP containment and requirements of the analysis tool was described. And physical models necessary for the hydrogen safety analysis code were listed. As a member of international collaborative project HYMERES for containment thermal hydraulics, KAERI is actively participating in an analytic working group. As an analysis tool for blind benchmarkes, the analysis code described in this paper was used. From the blind benchmark analyses, it was found that the code is very promising for hydrogen safety analysis. Currently, it is proposed to develop the code collaboratively in a hydrogen safety community based on an open-source strategy.

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

Atomic energy in Lithuania: Nuclear safety in 1996; Atomine energetika Lietuvoje: Branduoline sauga 1996 m

Energy Technology Data Exchange (ETDEWEB)

Vaishnys, P; Krenevichius, R; Alejev, A; Demchenko, M [State Nuclear Power Safety Inspectorate (VATESI), Vilnius (Lithuania); and others

1998-12-31

The first annual report of VATESI - Lithuanian Nuclear Safety Authority`s activity is presented and cover description of the main results in 1996. VATESI was established in 1991 recently after regaining of independence when Ignalina NPP come into jurisdiction of Republic of Lithuania. Since the establishment in six year period of operation VATESI developed its activities considerably and now is able to perform the functions of independent regulator of nuclear safety in Lithuania. The main fields of VATESI activities in 1996 were evaluation of the safety of Ignalina NPP, supervision of Ignalina NPP`s operational safety, control of the implementation of safety improvement program SIP-1 in Ignalina NPP, development of new regulatory legislation according to IAEA standards and practices, accountancy and control of nuclear materials, regulation of radioactive waste management. Detailed description of all these activities is provided in the report. In 1996 there were no safety significant events in Ignalina NPP according to the INES scale, only five events received qualification as level 1 events. 2 tabs.,10 figs.

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

Experience of TVSA fuel implementation at Kozloduy NPP

International Nuclear Information System (INIS)

Kamenov, K.; Kamenov, AI.; Hristov, D.

2011-01-01

The base design of the Russian fuel assemblies TVSA have been under operation at Kozloduy NPP WWER-1000 reactors since 2004. The old type fuel assemblies TVS-M were gradually substituted till 2008. The TVSA assembly distinguishes itself with much stronger construction. As a burnable absorber it has a mixture of uranium and uniformly distributed Gd in 6 or more fuel rods. This enables to increase the safety and effectiveness of fuel cycles. The experience gained during TVSA fuel implementation on units 5 and 6 and KASKAD code package validation was presented at the eightieth International conference on WWER 'Fuel performance, modelling and experimental support in 2009'. Additional information about TVSA fuel implementation at Kozloduy NPP WWER-1000 units in a 4-year fuel cycle with 42 and 48 fresh fuel assemblies reloading scheme is presented in the paper. (Authors)

30-th anniversary of nuclear power engineering in the USSR. Safety problems of nuclear engineering

International Nuclear Information System (INIS)

Kulov, E.V.; Sidorenko, V.A.; Kovalevich, O.M.

1984-01-01

The problems of safety and protection of personnel, population and environments from the NPP radiation effect in the Soviet Union in cases of possible accidents and in the course of normal operation are discussed. The problem of safety assurance includes scientific technological and organizational aspects. The necessity of accounting the ''national factor'' in the form of a possible NPP location, existing legislation experience gained, etc. is noted. The main trends in ensuring safety are a high quality of equipment manufacturing and assembly, quality control of equipment at all stages of operation, development and realization of efficient protective measures and devices, assurance of safety precautions at all stages of designing equipment production, NPP operation and maintenance, functioning of the system of state control of NPP safety assurance. The skill and training of personnel are necessary prerequisite for ensuring safety of nuclear power plants

The Improved Methods of Critical Component Classification for the SSCs of New NPP

International Nuclear Information System (INIS)

Lee, Sang Dae; Yeom, Dong Un; Hyun, Jin Woo

2010-01-01

Functional Importance Determination (FID) process classifies the components of a plant into four groups: Critical A, Critical B, Minor and No Impact. The output of FID can be used as the decision-making tool for maintenance work priority and the input data for preventive maintenance implementation. FID applied to new Nuclear Power Plant (NPP) can be accomplished by utilizing the function analysis results and safety significance determination results of Maintenance Rule (MR) program. Using Shin-Kori NPP as an example, this paper proposes the advanced critical component classification methods for FID utilizing MR scoping results

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)

Analysis of a buried pipeline at WWER-440/213 Paks NPP

International Nuclear Information System (INIS)

1999-01-01

According to regulations, all safety structures of a NPP must be designed to withstand loads induced by earthquakes. This applies to safety related underground piping systems. These structures are typically embedded in about 2-3 m of layered soil, and sometimes protected by concrete slabs resting on the ground surface. A rigorous solution for the dynamic response of such a structure would require accounting for nonlinear and three dimensional effects. A non-linear analysis is possible by using specialized computer codes and material models to account for non-linear behaviour of the soil. The aim of this paper was to reanalyze the pipeline of Paks NPP in order to demonstrate its dynamic behaviour interacting with the soil and the connected buildings as well as to determine the dynamic responses and the stresses in typical regions of the pipeline. To perform the numerical analysis a three dimensional finite element code (SSASSI/S) based on 'flexible volume method' was applied

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

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)

'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.)

Digital Components in Swedish NPP Power Systems

International Nuclear Information System (INIS)

Karlsson, Mattias; Eriksson, Tage

2015-01-01

Swedish nuclear power plants have over the last 20 years of operation modernised or exchanged several systems and components of the electrical power system. Within these works, new components based on digital technology have been employed in order to realize functionality that was previously achieved by using electro-mechanical or analogue technology. Components and systems such as relay protection, rectifiers, inverters, variable speed drives and diesel-generator sets are today equipped with digital components. Several of the systems and components fulfil functions with a safety-role in the NPP. Recently, however, a number of incidents have occurred which highlight deficiencies in the design or HMI of the equipment, which warrants questions whether there are generic problems with some applications of digital components that needs to be addressed. The use of digital components has presented cost effective solutions, or even the only available solution on the market enabling a modernisation. The vast majority of systems using digital components have been operating without problems and often contribute to improved safety but the challenge of non-detectable, or non-identifiable, failure modes remain. In this paper, the extent to which digital components are used in Swedish NPP power systems will be presented including a description of typical applications. Based on data from maintenance records and fault reports, as well as interviews with designers and maintenance personnel, the main areas where problems have been encountered and where possible risks have been identified will be described. The paper intends to investigate any 'tell-tales' that could give signals of unwanted behaviour. Furthermore, particular benefits experienced by using digital components will be highlighted. The paper will also discuss the safety relevance of these findings and suggest measures to improve safety in the application of digital components in power systems. (authors)

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)

Application of new technologies for the nondestructive testing equipment Novovoronezh NPP-2 and Leningrad 2

International Nuclear Information System (INIS)

Nichev, V.; Cvitanović, M.; Nadinic, B.

2016-01-01

This presentation demonstrates the latest technology and means of nondestructive testing of equipment of reactor VVER-1200, realized at the Novovoronezh NPP-2 and Leningrad NPP-2. The developments are based on a contract between ATOMKOMPLEKT with HRID, Croatia as a designer and contractor and and Rosatom. The activities of the presented company and IQC are associated with the qualification of methodologies control of components important to safety as required by the European methodology and requirements of Russian legislation

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

Performance indicator system with application to NPP management

International Nuclear Information System (INIS)

Gomez, J.; Roldan, J.

2001-01-01

The objective of the paper is to present the work that is being conducted in the scope of a research project between Cofrentes NPP and the polytechnic university of Valencia aimed to the development and implementation of a performance indicators system to support plant management. In developing this system, attention is being paid to the areas of safety, production and dependability. The first step in the project was the development of the performance indicator system (PIS), in order to help in assessing the effectiveness of the different activities in plant (i.e. maintenance, inspections, tests, etc.). It is suggested establishing the operational indicators set in 3 levels. The lowest level concerns indicators monitoring performance and maintenance characteristics of components. The next one involves a subset of indicators placed at system level with a similar goal. And finally, the highest level summarizes the impact of the global policy in the whole plant from safety and performance point of view. The definition of an indicator should comprise, at least, the following items: indicator's name, performance area, definition and data needed. A strategy should define what, when and how indicators have to be evaluated, analyzed and reported. This article gives an example application of the methodology at the Cofrentes NPP, collective dose as safety indicator, power production as production indicator and the number of work orders as maintenance indicator are considered and their time evolution is given. (A.C.)

Response and capacity evaluation of unit 1-2, Kozloduy NPP

International Nuclear Information System (INIS)

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

1993-01-01

Investigation described in this presentation was performed within the WANO Program for seismic safety assessment of Kozloduy NPP. The investigation is imposed by the necessity of seismic upgrading of the structures and equipment of the plant for the new design basic earthquake. Term of reference for this study was elaborated by experts of IAEA

Features of the Kozloduy NPP management system