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Application of the Code of Conduct in Polish research reactor MARIA
Regional Meeting on Application of the Code of Conduct on the Safety of Research Reactors, Lisbon, Portugal, 2-6 November 2015 Application of the Code of Conduct in Polish research reactor MARIA Andrzej Gołąb National Centre for Nuclear Research Otwock-Świerk, Poland
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Content Short presentation of High Flux Research Reactor MARIA
The last safety review events of research reactor Maria PERIODIC SAFETY REVIEW OF REACTOR MARIA Safety Analysis of the MARIA reactor 4.1. Decrease of the core cooling capability by means of the fuel channel circuit and pool cooling circuit 4.2. Deterioration of the cooling feasibility by the secondary circuit 4.3. Insertion of positive reactivity and power fluctuation 4.4. Failures of the core structural components or experimental equipment 4.5. Accidents induced by external events 4.6. Beyond design accidents 4.7. Accidents induced by internal events Ageing management Decommissioning Plan Emergency Plan for reactor Maria Conclusion
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1. Short presentation of High Flux Research Reactor MARIA
Designed and constructed by Polish industry First criticality reached in December 1974 1985 ÷ 1991 – modernization period: Put again into operation in 1992 3
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General characteristics of MARIA reactor
Nominal power 30MW Maximum thermal neutron flux: in fuel in beryllium 2.5 · 1018 n/m2s 4.0 · 1018 n/m2s Moderator water and beryllium Reflector graphite (blocks in Al cans) and water Fuel element: Material and enrichment shape overall dimensions dispersion U3Si2 in Al. - 19,75% U-235 5 concentric tubes 100 cm length Primary fuel cooling system: type of fuel channel pressure range temperature, core inlet (outlet), water flow rate: through channel total Field tube 0.8 ÷ 1.8 Mpa 50 (100) ºC 25 m3/h or 30 m3/h 550 ÷ 650 m3/h Primary pool cooling system: pressure temperature: at core matrix inlet at core matrix outlet Water flow rate Atmospheric 40 ºC 50 ºC 1400 m3/h
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The main areas of reactor application are:
production of radioisotopes irradiation of uranium plates for Mo-99 production testing of fuel and structural materials for nuclear power engineering neutron radiography neutron activation analysis neutron transmutation doping research in neutron and condensed matter physics training
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reactor relicensing – March 2015 updating following documents:
2. The last safety review events of research reactor Maria: INSARR mission – April 2014 reactor relicensing – March 2015 updating following documents: safety analysis report, emergency preparedness plan, preliminary decommissioning plan, ageing management programme, radiation protection programme, quality assurance programme, classification of system and components important to safety.
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3. PERIODIC SAFETY REVIEW OF REACTOR MARIA
On the base of Atomic Law (art. 37e) NCBJ is obligated to carried out safety review of reactor Maria every 4 years. Plan of safety review has to be approved by President of National Atomic Energy Agency (Regulatory Body). Plan has to be submitted to the President of NAEA 6 months before beginning of the safety review. Report of safety review has to be submitted to the President of NAEA, not later than 3 months after carrying out the safety review and has to be approved.
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4. Safety Analysis of the MARIA reactor
Classification of abnormal events regarded in Safety Analyses of MARIA reactor is carrying out on the base of Atomic Law and the Regulation of The Council of Ministers of 31 August 2012. Adapting recommendation of IAEA, contained in Safety Requirements NS-R-4, to specific of reactor Maria the following initiating events were discussed:
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Coolant flow vanishing in the fuel channel circuit
4.1. Decrease of the core cooling capability by means of the fuel channel circuit and pool cooling circuit Coolant flow vanishing in the fuel channel circuit Blocking of coolant flow in a fuel channel Bypassing of flow through the fuel element Loss of the tightness in the fuel element cooling circuit Decline of water flow in the pool cooling system Water escape from reactor pool cooling circuit Bypassing of coolant flow rate in the pool through the slots in the core matrix
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Reactivity disturbances when the reactor is scrammed
4.2. Deterioration of the cooling feasibility by the secondary circuit 4.3. Insertion of positive reactivity and power fluctuation Reactivity disturbances when the reactor is scrammed Reactivity disturbances when the reactor start-up Reactivity disturbances during the operation on nominal power Reactivity disturbances during fuel storing
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4.4. Failures of the core structural components or experimental equipment
Mechanical failure of the fuel element Failures of core structural components Loss of air-tightness of the can containing the target material 4.5. Accidents induced by external events Accidents associated with power supply system Earthquake impact on reactivity disturbances Fall of an airplane on the reactor External flood
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4.6. Beyond design accidents
Breach of the main piping or the fuel channel Partial melting of reactor core 4.7. Accidents induced by internal events Internal fire Internal flood Generally deterministic approach is applied in reactor safety analysis. In the case of mechanical failure of the fuel element and fall of an airplane on the reactor safety analysis are completed by probabilistic approach.
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5. Ageing management Ageing monitoring of reactor systems and components is carried out on the base of procedure: „Ageing management of reactor Maria” No. 03-ZR-15 – updated this year. This procedure contains the list of reactor systems and components which are surveyed due to ageing and methods and intervals of examination are presented. To keep the reactor technical state in high level, assuring its safety and disposebility continuous process of reactor modernization is carried out.
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In the last few years the following important modernization tasks were performed:
implementation of a new neutron measurement lines based on Hartman-Brown’s instrumentation, implementation of a new instrumentation for controlling thermo-hydraulics reactor parameters such as: temperatures, flow-rates, pressure, modernization of radiation protection systems, a new system is based on „intelligent” Eberline detectors, modernization of a fuel elements integrity detection system, replacement of heat exchangers internals,
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implementation of a new visualisation and recording system (SAREMA),
modernization of a preparatory station for secondary circuit water supplying system, implementation of a new visualisation and recording system (SAREMA), replacement of batteries in emergency electrical supplying system, Replacement of primary cooling system main pumps. Besides, the continuous process of ageing control of following important reactor components is being carried out, i.e.: reactor core beryllium blocks, reactor reflector graphite blocks, pipe-lines of primary cooling systems. 15
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6. Decommissioning Plan Preliminary Decommissioning Plan has been elaborated as the consequence of implementation of Code of Conduct. This plan contains: Identification of radioactive contaminated materials Scope of dismantling activities Description of dismantling technology Techniques of dismantling Techniques of decontamination Waste management Rules of organization 16
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The main principles to be undertaken are:
to perform partial decommissioning (safe enclosure) decommissioning will be carried out as soon as possible after reactor shut-down decommissioning will be carried out by reactor operational personnel
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7. Emergency Plan for reactor Maria
The reactor facility is a part of the National Centre for Nuclear Research (NCBJ). It is physically separated by special protection system. General Director of NCBJ appoints the Deputy Director of NCBJ for Nuclear Safety and Radiation Protection (DNSRP) and establishes Emergency Service for Nuclear Centre. The Service is aimed to prevent hazardous events, supervise proper preparation of the emergency measures and to mitigate consequences of the accidents. The Service pursues continuous supervision of the condition of the NCBJ facilities.
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The Emergency Service operates closely with public organizations established to ensure safety. These include: Centre for Radiation Events of National Atomic Agency, Central Laboratory for Radiation Protection, State Fire Brigade, Health Centre in Otwock, Crisis Management Centre in Otwock.
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EMERGENCY CONSULTANTS
GENERAL DIRECTOR EMERGENCY MANAGER of Nuclear Centre EMERGENCY DISPATCHER EMERGENCY CONSULTANTS Chief of Internal Security Service Guard Emergency Managers of Nuclear Facilities Emergency Groups of Facilities Deputy for Radiation Protection of Nuclear Centre Deputy for Radiation Protection of POLATOM Deputy for Technical Measures Special Technical Teams Maintenance Personnel Fig. 8. An organizational chart of Emergency Service for Nuclear Centre
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The organization of reactor MARIA operational staff
Reactor Emergency Manager (REM) appointed directly by General Director of NCBJ. Reactor Emergency Group to be called Emergency Service for the reactor MARIA facility. The group consists of selected and trained members of the operational staff. REM is the manager of the reactor MARIA operational staff. The following is associated with the position of REM: deputy manager of the MARIA reactor and manager of the reactor shift.
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Emergency levels Emergency Plan identifies the following levels of hazard: internal – in the reactor site, local – in the area of the Nuclear Centre, public – outside the Nuclear Centre. The limits are established in the Emergency Plan which are the basis for making decision to start the emergency actions.
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Practical training includes:
emergency exercises including cooperation of involved units (Reactor Emergency Group, Emergency Dispatcher of Nuclear Centre, technical services, internal protection services, external units) emergency equipment usage
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8. Conclusion Significant improvements in application of the Code of Conduct are following: Assuring of full independence of Regulatory Body and Operating Organization Elaboration of succession plan for the personnel to assure adequate human resources needed for safe reactor operation Reorganization of radiological zoning inside reactor building Development of ageing management programme Elaboration of classification of system and components important to safety Development of maintenance program Modification of reactor ventilation system in order to minimize radioactive releases in accident conditions 24 24
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New activities in implementation of the Code of Conduct are following:
Development of procedure for the core configuration change Assuring the independence of reactor radiation protection officer from the reliance of reactor manager Improvement of reactor protection system to assure redundancy of some safety channels
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Thank you for your attention
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