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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 ResearchOtwock-Świerk, Poland

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Content1. SHORT PRESENTATION OF HIGH FLUX RESEARCH REACTOR MARIA

2. THE LAST SAFETY REVIEW EVENTS OF RESEARCH REACTOR MARIA

3. PERIODIC SAFETY REVIEW OF REACTOR MARIA

4. SAFETY ANALYSIS OF THE MARIA REACTOR 4.1. Decrease of the core cooling capability by means of the fuel channel

circuit and pool cooling circuit4.2. Deterioration of the cooling feasibility by the secondary circuit4.3. Insertion of positive reactivity and power fluctuation 4.4. Failures of the core structural components

or experimental equipment 4.5. Accidents induced by external events4.6. Beyond design accidents4.7. Accidents induced by internal events

5. AGEING MANAGEMENT

6. DECOMMISSIONING PLAN

7. EMERGENCY PLAN FOR REACTOR MARIA

8. CONCLUSION

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Designed and constructed by Polish industry First criticality reached in December 1974 1985 ÷ 1991 – modernization period: Put again into operation in 1992

1. SHORT PRESENTATION OF HIGH FLUX RESEARCH REACTOR MARIA

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Nominal power 30MW

Maximum thermal neutron flux:in fuelin beryllium

2.5 · 1018 n/m2s4.0 · 1018 n/m2s

Moderator water and beryllium

Reflector graphite (blocks in Al cans) and waterFuel element:Material and enrichment

shapeoverall dimensions

dispersion U3Si2 in Al. - 19,75% U-235

5 concentric tubes100 cm length

Primary fuel cooling system:type of fuel channelpressure rangetemperature, core inlet (outlet), water flow rate:through channeltotal

Field tube0.8 ÷ 1.8 Mpa50 (100) ºC

25 m3/h or 30 m3/h550 ÷ 650 m3/h

Primary pool cooling system:pressuretemperature:at core matrix inletat core matrix outletWater flow rate

Atmospheric

40 ºC50 ºC

1400 m3/h

General characteristics of MARIA reactor

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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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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.

2. THE LAST SAFETY REVIEW EVENTS OF RESEARCH REACTOR MARIA:

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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

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

4.1. Decrease of the core cooling capability by means of the fuel channel circuit and pool cooling circuit

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• 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

4.2. Deterioration of the cooling feasibility by the secondary circuit

4.3. Insertion of positive reactivity and power fluctuation

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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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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.

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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

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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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GENERAL DIRECTOR

EMERGENCY MANAGERof Nuclear Centre

EMERGENCY DISPATCHERof Nuclear Centre

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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Significant improvements in application of the Code of Conduct are following:

1. Assuring of full independence of Regulatory Body and Operating Organization

2. Elaboration of succession plan for the personnel to assure adequate human resources needed for safe reactor operation

3. Reorganization of radiological zoning inside reactor building

4. Development of ageing management programme

5. Elaboration of classification of system and components important to safety

6. Development of maintenance program

7. Modification of reactor ventilation system in order to minimize radioactive releases in accident conditions

8. CONCLUSION

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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

New activities in implementation of the Code of Conduct are following:

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Thank you for your attention