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IAEA-TCR-07264ORIGINAL: English
INTERNATIONAL ATOMIC ENERGY AGENCY
REPORT OF THE
INTEGRATED SAFETY ASSESSMENT OF RESEARCH REACTORS
(INSARR) MISSION
TO THE
SLOVENIA TRIGA MARK- IIRESEARCH REACTOR
Ljubljana, Slovenia12-16 November 2012
Conducted under IAEA Technical Co-operation Project: SLO1006
DEPARTMENT OF TECHNICALCOOPERATION
Division for Europe
DEPARTMENT OF NUCLEAR SAFETY AND SECURITY
Division of Nuclear Installation Safety
INTERNATIONAL ATOMIC ENERGY AGENCYORIGINAL: ENGLISH
Organized by: IAEA Technical Cooperation Project: SLO1006
Conducted by: Mr A. M. Shokr (IAEA/NSNI, SH-RRSS- Team Leader)Mr. H. Abou Yehia (IAEA/NSNI, Deputy Team Leader)Mr. H. Hirshfeld (IRR-1, IAEC, Israel)Ms. V. Garea (INVAP, Argentina)Mr. A. Jraut (CNSTEN, Morocco)
CONTENTS
1 INTRODUCTION............................................................................................................. 1
1.1. BACKGROUND............................................................................................................... 11.1.1 Short description of the facility.......................................................................... 11.1.2 History of the facility including the modifications ............................................ 11.1.3 Utilization programme ....................................................................................... 1
1.2. OBJECTIVE AND SCOPE OF THE MISSION .............................................................. 21.3. BASIS AND REFERENCES FOR THE REVIEW.......................................................... 21.4. CONDUCT OF THE MISSION ....................................................................................... 4
2 CONCLUSION AND SUMMARY OF MAIN RECOMMENDATIONS ...................... 6
3 RESULTS OF THE MISSION ......................................................................................... 8
3.1. FACILITY WALKTHROUGH ........................................................................................ 83.2. OPERATING ORGANIZATION AND REACTOR MANAGEMENT.......................... 93.3. SAFETY COMMITTEE................................................................................................. 103.4. TRAINING AND QUALIFICATION............................................................................ 113.5. SAFETY ANALYSIS REPORT..................................................................................... 123.6. SAFETY ANALYSIS ..................................................................................................... 123.7. OPERATIONAL LIMITS AND CONDITIONS ........................................................... 133.8. CONDUCT OF OPERATIONS...................................................................................... 133.9. MAINTENANCE PERIODIC TESTING AND INSPECTION .................................... 143.10. UTILIZATION AND MODIFICATIONS ..................................................................... 153.11. RADIATION PROTECTION AND RADIOACTIVE WASTE MANAGEMENT ...... 163.12. EMERGENCY PLANNING........................................................................................... 173.13. QUALITY ASSURANCE .............................................................................................. 183.14. DECOMMISSIONING................................................................................................... 183.15. REGULATORY SUPERVISION................................................................................... 18
ANNEX 1: AGENDA.............................................................................................................. 20
ANNEX 2: LIST OF PARTICIPANTS................................................................................... 22
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1 INTRODUCTION
1.1 BACKGROUND
1.1.1 Short description of the facility
The Slovenia research reactor is a TRIGA MARK-II research reactor with a nominal steady state power of 250 KW, and of a maximum pulse power of 1000 MW and operated by the Jozef Stefan Institute (IJS), Ljubljana, Slovenia. The reactor is subject to Project and Supply Agreement, under which the Agency’s Safety Standards apply to the reactor operation.
The TRIGA MARK II reactor core is housed in an aluminum vessel of 6.25 m height and 2 m diameter. The reactor fuel material is harmonized mixture of 20 % enriched uranium and zirconium hydride. The atomic ratio between hydrogen and zirconium in the zirconium is 1.6. The total number of fuel elements in the core is currently 60 fuel elements. The reactor is cooled and moderated by light water and with graphite reflector. The layout allows for access to the core from the top of the reactor platform, thus greatly simplifying handling of fuel and core component including experimental devices.
There are six experimental channels installed in the reactor: large and small thermal column, two tangential channels, one to the core, the other to the reflector, and two radial channels, one to the core and one to the reflector. The channels are stepped tubing, installed in the concrete shield. The small thermal column leads into a larger protected experimental chamber named the dry cell. The reactor is situated in an octagonal reactor building with a volume of 6000 cubic meters. Ventilation and heating are carried out by a ventilation system, which also keeps the pressure within the reactor building below the atmospheric pressure.
The IJS has recently initiated a feasibility study for the installation of a thermal neutron-driven 14 Mev converter. A national Technical Cooperation project (TC-SLO1006) was initiated in cooperation with the IAEA in 2012 to support the IJS on performing such a feasibility study. This Integrated Safety Assessment for Research Reactor (INSARR) mission is conducted under this TC project.
1.1.2 History of the facility including the modifications
The reactor went critical for the first time on 31 May 1966. Since that time several modifications and refurbishment activities have been performed, including use of LEU fuel. These activities also included modernization of the Instrumentation and Control (I&C) system, and renovation of the reactor core components and the control rod drive mechanisms.
In 1999, the reactor was reconstructed and upgraded to pulse operation mode. The spent fuel was shipped back to the countries of origin, including shipment of spent fuel to USA (219 spent fuel elements), and France (10 fresh fuel elements and 600 Kg ADU yellow cake in 2007).
1.1.3 Utilization programme
The reactor is primarily used for research and development in physics and engineering including radiation hardness studies, neutron radiography, production of trace elements,
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education and training, and for neutron activation analysis. The reactor is effectively used for education and training nuclear engineering and radiation protection courses. It is used for training of future nuclear power reactor operators in Slovenia and frequently used to train fellows under different IAEA programmes on research reactor safety and operation. The current operating schedule of the reactor is about 700 hours per year. This schedule may vary depending on user’s requests.
1.2 OBJECTIVE AND SCOPE OF THE MISSION
The objective of the mission was to review the safety of the TRIGA MARK II reactor of the IJS, Slovenia, through the IAEA safety review Service INSARR, and to provide recommendations and suggestions for the reactor safety improvement. The review covered all the safety aspects of the reactor operation including:
1. Regulatory supervision;2. Operating organization and reactor management;3. Safety committee; 4. Training and qualification of operating personnel; 5. Safety Analysis Report (SAR);6. Safety analysis;7. Operational Limits and Conditions (OLCs), 8. Conduct of operations; 9. Maintenance, periodic testing and inspection; 10. Utilization and modifications; 11. Operational radiation protection and waste management programme; 12. Emergency planning;13. Quality assurance programme;14. Decommissioning plan.
1.3 BASIS AND REFERENCES FOR THE REVIEW
The following documents were used as reference for the review:
1) Guidelines for the Review of Research Reactor Safety, IAEA Service Series No.1, 1997;
2) Safety of Research Reactors, IAEA SS NS-R-4, IAEA, 2005;
3) Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, IAEA SS No. 35-G1, IAEA, 1994;
4) Safety in the Utilization and Modification of Research Reactors, IAEA Safety Guide SSG-23, IAEA, 2012;
5) Maintenance, Periodic Testing, and Inspection of Research Reactors, IAEA Safety Guide SG NS-G-4.2, IAEA, 2006;
6) Core Management and Fuel Handling for Research Reactors, IAEA Safety Guide NS-G-4.3, IAEA, 2008;
7) Operational Limits and Conditions and Operating Procedures for Research Reactors, IAEA Safety Guide NS-G-4.4, IAEA, 2008;
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8) Operating Organization and the Recruitment, Training and Qualification of Personnel for Research Reactors, IAEA Safety Guide NS-G-4.5, IAEA 2009;
9) Radiation Protection and Radioactive Waste Management in the Design and Operation of Research Reactors, IAEA Safety Guide NS-G-4.6, IAEA, 2009;
10) Ageing Management for Research Reactors, IAEA Safety Guide SSG-10, IAEA, 2010;
11) Safety Analysis Report (SAR) of the TRIGA MARK II, IJS-DP-5823, November 2009;
12) Set of national regulatory requirements for nuclear and radiation safety, (established during the period 2003-2013);
13) Rules on radiation and nuclear safety, JVS, No 29/2009;
14) Radiological protection programme for the TRIGA MARK II, IJS-DP-VPIS-100-01, 2012;
15) ALARA programme at IJS, SVPIS-R-DN-17;
16) Waste management procedures, SVIPS-R-DN-32;
17) Radioactive waste and spent fuel management programme, RIC-PrRAO-1;
18) Removal of material from controlled areas, SVPIS-R-DN-09;
19) Radiological survey of controlled area in TRIGA MARK II reactor, SVPIS-R-DN-26;
20) Operational waste management procedures, SVPIS-R-DN-32;
21) Reconstruction and modification of TRIGA MARK II reactor for pulse operation mode, IJS-DP-5762, 1990;
22) Reactor operating procedures for TRIGA MARK II reactor, IJS-DP-6016;
23) Training programme for the TRIGA reactor operators, IJS-DP-9296, Rev. 1, 2006;
24) Inspection plan of the TRIGA MARK II reactor, IJS-DP-6471, Rev 1, 1992;
25) Procedures to conduct activity at the TRIGA MARK II reactor, RIC-POG-DE-2;
26) Emergency plan and response (NUID-RIC-01), U1-QA-211, 2011;
27) Report on emergency exercise in October 2012, UNID-ZAP-1, 2012;
28) Decommissioning programme for the TRIGA MARK II reactor, IJS-DP-9849, 2007.
As they are written in Slovenian language, an English summary of the reactor safety documents was submitted to the review team before the mission. In addition to the documents mentioned above, several operation and maintenance documents and records were also reviewed during the mission, with direct translation made by the reactor technical staff from Slovenian to English.
The reactor management and the operating personnel made presentations on each of the review areas, mentioned above in Section 1.2. In addition, Mr T. Nemec of the Slovenia Nuclear Safety Administration (SNSA) provided a presentation on the licensing process and regulatory supervision of research reactors in Slovenia. These presentations provided an
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overview of the safety status of the relevant review areas, and were discussed in the frame of the mission activities.
1.4 CONDUCT OF THE MISSION
The mission was conducted by a team consisted of two IAEA staff members (Mr A. M. Shokr, SH-RRSS/NSNI, Team Leader), and Mr H. Abou Yehia (RRSS/NSNI, Deputy Team Leader), and three external experts; Mr H. Hirshfeld (Reactor Manager, IRR-1, Soreq Nuclear Research Centre, Israel Atomic Energy Commission), Ms V. Garea (Head of Safety Assessment and Environmental Impact Department, INVAP, Argentina), and Mr A. Jraut (Head of Safety Analysis Unit, CNESTEN, Morocco).
The objectives and scope of the mission were reviewed in the opening meeting with the participation of Mr J. Lenarcic, Director of the IJS. Representatives of the SNSA participated in the activities of the mission. Mr Andrej Stritar, SNSA Director attended some working sessions and participated in the exit meeting. The agenda of the mission and the list of participants are provided in Annex I and Annex II, respectively.
The mission conclusions and recommendations were discussed with the reactor operating personnel and management in the exit meeting, with general agreement on the IAEA recommendations and suggestions.
The review was conducted based on the following actions:
a) Review of safety and operating documents;
b) Discussions with the reactor management and technical staff;
c) Discussions between the team members; and
d) Field reviews, including walkthrough the reactor facilities.
Review criteria:
The review process compared the observations and findings of the team with the IAEA Safety Standards. This comparison resulted in recommendations, suggestions comments and good practices, presented to the operating organization by the team as a whole, in accordance to the following definitions:
RecommendationIs a team advice to improve the safety, it will be reviewed during the follow up and it is based on IAEA Safety Standards. It will focus on WHAT to do. However, under Comments approaches on the HOW can be mentioned.
Suggestion Is a team proposal in conjunction either with a recommendation or may stand on its own. It may indirectly contribute to improvements the safety but is primarily intended to enhance performance.
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Good Practice It is a proven performance, activity, or use of equipment, which the team considers to be markedly superior to that observed elsewhere. It should have broad application to other facilities.
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2 CONCLUSION AND SUMMARY OF MAIN RECOMMENDATIONS
The implementation of the mission was done in good conditions where the openness and motivation of the reactor management and operating personnel and their good preparation for the mission should be highlighted. The team also recognized and appreciated the commitment of the IJS management to improve the reactor safety.
The team noted with satisfaction the good practices and efforts to enhancing safety in several areas including in particular establishment of a comprehensive training and qualification programme and performance of Periodic Safety Review (PSR). The team made recommendations and suggestions to further enhancing the safety of the reactor. The team recommendations and suggestions covered mainly the organizational aspects, safety analysis and safety documents, and technical modifications to the facility.
The main recommendations made by the mission team are the following:
Organization:
The Terms of Reference of the Safety Committee (SC) should be revised to include the advisory role of the Committee to the IJS director and to ensure consistency with the IAEA Safety Standards and international good practices concerning the items to be reviewed by the committee.
The roles and responsibilities for safety of the Head of Reactor Infrastructure Centre, reactor manager and the reactor operating staff should be defined by the IJS in a formal document.
The duties and responsibilities of the Technical Manager (reactor manager) for reactor operation should be covered by a full time position in the IJS organization and the necessary funds should be ensured by the Government for this purpose.
Safety analysis and safety documents:
The technical content of the revised SAR should be in accordance with the IAEA Safety Standards. SAR should include a revised version of safety analysis which should cover selection of Postulated Initiating Events (PIEs), description of event sequence and comparison against acceptance criteria. The OLCs and emergency plan should be revised to reflect the results of the safety analysis.
The operating procedures should be completed (revised) to cover all the operations of safety significance and to ensure realization of the OLCs, including the procedures for fuel loading into reactor core.
Technical modifications to the facility:
The fire hazard analysis should be completed as soon as possible and, accordingly, fire detectors (or automatic extinguishers as necessary) should be installed at all reactor
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areas with potential fire hazard. To reduce the fire load, all non-used inflammable materials should be removed from the reactor building.
The emergency ventilation system should be equipped with charcoal filters. Periodic testing of the efficiency of the filtration system should be performed.
Monitoring of the reactor water leakage from reactor pool, beam tubes, primary pumps and heat exchangers should be improved by installation of adequate detection system.
Concerning the interaction with the regulatory body, the team noted that the regulatory supervision of the reactor is adequate and effective. The team suggested to include in the topics of future inspections the verification of application of OLCs, in particular those related to periodic testing.
Section 3 of this report provides detailed discussions on the safety issues identified by the team during the mission and the associated recommendations and suggestions. While it is not a review area that is covered by this mission, Section 3 of the report also includes some remarks on the reactor design, as it is closely related to (and has influence on) safety.
The reactor management is recommended to establish an action plan and a follow-up for implementation of the recommendations formulated during this mission. This plan should include the actions to be taken with the persons in charge and the associated deadlines.
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3 RESULTS OF THE MISSION
3.1 FACILITY WALKTHROUGH
The team visited the reactor on 12 November 2012 and performed a walkthrough the facility. During this technical visit, the reactor was in operation at 60 KW for training of university students. The visit familiarized the team members with the reactor facility and updated themon the safety status of the reactor systems and components. During this visit, the team discussed with the operating personnel different aspects of the safe operation of the reactor, including mainly operating procedures, operational radiation protection aspects, fire safety, and detection of water leakage from different systems and components. The following areas were visited:
Reactor control room; Reactor pool top; Reactor experimental area; Water treatment system and area of cooling systems components; Ventilation system area; Hot cells and chemical laboratories.
During this visit, the housekeeping was observed to be good. Some operating documents were available in the control room. The checklists of the reactor start-up, power operation, and shutting down were found to be complete as well as the records of the important operating parameters. The information reported in the operation logbook was found to be complete, and there was evidence of their review by the reactor manager. The operability check of the control room alarm lamps which was performed by the reactor operator based on a request from the team showed the proper functioning of these lamps. However, it was observed that control room console still include some pushbuttons (and associated indicators) related to components that have been already removed from the service. These pushbuttons and indicators need to be removed.
The team observed that the over-coats used in the controlled areas are also used in areas that are categorized from radiation protection point of view as supervised. No over-shoes were used in the controlled areas (see also Section 3.11).
During walkthrough the reactor facility, lack of fire detectors in most of areas within the reactor building was observed. The operating personnel mentioned that actions have been initiated to perform fire hazard analysis and fire protection system will be installed accordingly. Significant fire loads were observed in several locations within the reactor experimental hall (accumulated boxes and pallets, paraffin blocks inside wooden boxes).
During the walkthrough it was also realized that there is no charcoal filter in the emergency ventilation system. The operating personnel mentioned that the efficiency of the absolute filters associated with the ventilation system is not measured, and there are no requirements within the OLCs on the measurement of the filters' efficiency.
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The walkthrough showed also the need to improve the detection of possible water leakage from important components including primary pumps and beam tubes.
Recommendations:
R1) The reactor fire safety should be strongly improved. The fire hazard analysis should be completed as soon as possible and, accordingly, fire detectors (or automatic extinguishers as necessary) should be installed at all areas of the reactor building with potential fire risk including control room, reactor hall, experiments area, ventilation and primary pump areas, and electrical switchboard room. The fire safety programme should be included in a revised version of the SAR. All non-used items of potential fire risk should be removed from the reactor building.
R2) The emergency ventilation system should be equipped with charcoal filters. The specifications of the filters and criteria of verification of their efficiency should be established and included in the OLCs.
R3) Monitoring of reactor water leakage from reactor pool, beam tubes, primary pumps and heat exchangers should be improved by installation of adequate detection and monitoring system.
3.2 OPERATING ORGANIZATION AND REACTOR MANAGEMENT
The IJS has prime responsibility for safety. In this regard, the Director of IJS is the prime responsible for safe operation of the reactor. The Head of the Reactor Infrastructure Centre(RIC) is appointed by the Director of IJS to manage the safe operation of the reactor and the applied research activities as well as to manage the safe operation of the hot cells facility. The operation of the reactor is ensured by a technical manager (reactor manager) assisted by anoperation shift supervisors and reactor operators. An independent support in the field of radiation protection is provided to the reactor by the IJS Radiation Protection Unit.
The current administrative and organizational provisions for the reactor operation do not entail a full time position for the reactor manager. This situation is not in line with the IAEA Safety Standards and international practices.
Currently, in the IJS organization there is no clear and formal definition of the roles and responsibilities of the Head of RIC, the reactor manager and the reactor operating staff. Such definition is needed to ensure complementarities and avoid unnecessary overlapping of responsibilities for safety.
Financial resources for the safe operation of the reactor are provided by the Slovenian Government to the IJS. According to the IAEA Code of Conduct on the Safety of Research Reactors, the Government of Slovenia should provide necessary funds to ensure and maintain the safe operation of the reactor and to cover the cost of a full time position for the reactor manager.
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Recommendations:
R4) The roles and responsibilities of the Head of RIC, the Technical Manager (reactor manager) and operating staff, should be defined in a precise and formal manner. The duties and responsibilities of the reactor manager should be covered by a full time position in the IJSorganization. The definition of roles and responsibilities for ensuring safe operation of the TRIGA MARK II reactor should be integrated in the SAR and OLCs.
R5) Recommendation to the Government of Slovenia: Necessary funds should be ensured by the Government for enhancing safe operation of the reactor, which requires that the reactor manager ensures his duties in a full time position within the IJS.
3.3. SAFETY COMMITTEE
The Safety Committee (SC) is appointed by the Director of IJS for a fixed term of four years. It is composed of eight members including the chairman, five members not having responsibility for reactor operation, two members with direct responsibility for radiation protection and one member involved in the reactor operation.
The SC prepares minutes of its meetings and it is required to meet at least once per year. The minutes of each meeting are sent to IJS Director and Head of the RIC. The committee performs review and audits of activities related to the safety and operation of the reactor and to radiological safety in the hot cells. In particular, the SC reviews the modifications and experiments with safety significance, performs audits and prepares annual reports to IJS Director presenting the findings.
According to the IAEA Safety Standards, the SC has an advisory role for the benefit of the head of the operating organization. This aspect is not sufficiently highlighted in its current terms of references and it is not correctly reflected in the OLCs which indicates an authorization role.
In the current composition of the SC, its chairman is the Head of the Quality Assurance (QA) Unit in IJS which ensures an internal supervision of the QA activities related to the operation of the reactor. This situation presents a potential for conflict of interest between the two functions. In order to prevent such potential conflict, during the review of issues involving QA aspects there is a need that the SC Chairman assigns one member of this committee to chair the meeting when QA issues are addressed.
Recommendation:
R6) The terms of reference of the SC should be revised in accordance with the IAEA Safety Standards NS-R-4 to include in its functions the advisory role to the IJS Director, and to include the following in the list of items to be reviewed:
Proposed changes to the safety documents (SAR, OLCs, and emergency plan); Violations to the OLCs and to operating procedures of safety significance; Reports on routine releases of radioactive materials to the environment, and dose to
personnel and public;
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Events that are required to be reported or have been reported to regulatory body; Periodic reviews of the operational safety performance of the reactor (which could be
in the form of review of the annual report of reactor operation); Regulatory inspection reports.
The OLCs should be revised accordingly.
Suggestion:
S1) The revised terms of reference of the SC should indicate that the chair of this committee should delegate his role to another member when the committee reviews issues in which he has direct involvement.
3.4. TRAINING AND QUALIFICATION
A formal training and re-training programme is established for the reactor operating personnel. There are training and retraining programmes for the reactor operators, senior reactor operators and reactor manager. The programme covers basic theoretical knowledge and on-the-job training. The topics covered by the initial training programme include:
Principles of reactor physics, reactor technology, and reactor operation; Instrumentation and control of TRIGA MARK II reactor, and operating procedures; Reactor system description; Operational radiation protection and waste management protection; Reactor safety documents; Administrative requirements; Handling of fuel and radioactive materials; Training on emergency; Regulatory requirements for nuclear and radiation safety.
There are adequate training materials that cover the topics of the training programme. The training is provided on the basis of class-room presentations, self-study, and on-the-job training under the supervision of senior staff.
In accordance with the national regulatory requirements, a formal process is established for the licensing of the reactor operators. The licensing exam is administrated by the regulatory body. The reactor manager is part of the examination committee. The license should be renewed every four years on the basis of a formal retraining programme. The retraining programme for the reactor manager and the senior operators include their participation in regional and international forums, meetings and international activities on research reactors.
The operators and maintenance personnel are trained on operational radiological protection aspects. Reactor users are receiving short training on radiation protection administrativeprocedures and emergency plan of the reactor.
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Good Practice:
GP1) The participation of the reactor manager and rector senior operators in the regional and international forums/groups and their participation in international activities on research reactor as a part of their re-training programme is considered as a good practice.
3.5. SAFETY ANALYSIS REPORT
The latest version of the SAR was developed in 2009. The format of the SAR is different than the one recommended by the IAEA Safety Standards (IAEA SS No. 35-G1). The technical contents of the current version of SAR do not cover several areas that are important to safety including:
History of the reactor facility; Reactor safety objectives; Design requirements including safety classification of Structures, Systems, and
Components (SSCs); Material specifications; Utilization, including safety of experiments and utilization programme; Emergency plan; Significant amount of information related to safety analysis and OLCs.
The reactor operating personnel mentioned that a revised version of SAR is under development in accordance with the recently issued regulatory requirements.
Recommendation:
R7) The technical contents of the revised version of the SAR should be in accordance with the international practices established by the IAEA Safety Standards SS No. 35-G1. The contents of SAR should provide the reviewer adequate information to assess the safety of the reactor facility. The team comments (observations and recommendations) presented in different Sections of this mission report should be resolved in the revised version of the SAR including, among others, revised safety analysis, OLCs, and safety of experiments.
3.6. SAFETY ANALYSIS
The safety analysis presented in the current version of the SAR is not adequate. It presents the consequences of some events that involve radiation exposure which may be suitable for emergency planning purposes. The events considered are failure of fuel element in air, damage to fuel element cladding, accidental increase of reactivity, and Loss of Coolant (LOCA). The analysis is limited to the consequences in terms of dose to the public (and to operator in some cases).
It is not clear how the Postulated Initiating Events (PIEs) were identified and selected. The events' sequences are not described and there is no comparison of the safety analysis results against acceptance criteria. The team discussed with the reactor staff the steps and methods of
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performing safety analysis for research reactors following the international good practices and IAEA Safety Standards (NS-R-4, SS No. 35-G1, and SRS No. 55).
The discussions covered list of PIEs that may be applicable to the TRIGA MARK II reactor, description of event sequence and calculations of the event consequences.
Recommendation:
R8) The safety analysis should be developed and included in the revised version of the SAR. The list of PIEs included in the Appendix of the IAEA Safety Standards NS-R-4 could be taken as a basis for the development of list of the PIEs specific for the TRIGA MARK II reactor. The analysis should include description of the event sequence, event consequencesand comparison against acceptance criteria. The results of the safety analysis should be reflected in the establishment of the OLCs.
3.7. OPERATIONAL LIMITS AND CONDITIONS
The information provided in the SAR does not make distinction between different groups of the OLCs: Safety Limits; Safety System Settings (SSS), Limiting Conditions for Safe Operation (LCOs); Surveillance requirements (or periodic testing); and Administrative Requirements. Even for the few specifications provided in the current version of SAR, it is not clear which specification belongs to which group of OLCs. These specifications are not presented in terms of the internationally used attributes (objective, basis, and applicability of the specification) that are recommended by the IAEA Safety Standard NS-G-4.4. There is no chapter on the LCOs. The current version of SAR does not show how the results of the safety analysis were used to establish the OLCs.
The team explained the IAEA safety requirements and guidance on the establishment of OLCs and provided examples for application to the case of the TRIGA MARK II reactor.
Recommendation:
R9) The operating organization is recommended to establish, and include in the SAR, a complete set of the OLCs on the basis of the results of a revised safety analysis. Appendix I of the IAEA Safety Guide NS-G-4.4 could be considered as guidance to establish LCOs applicable to the TRIGA MARK II reactor. Annex II of the IAEA Safety Guide NS-G-4.2 provides examples for periodic testing activities and could be used for establishing the surveillance requirements of the TRIGA MARK II reactor.
3.8. CONDUCT OF OPERATIONS
Operation of the reactor is covered by operating procedures. The Quality Assurance (QA)programme includes list of the operating procedures and requires review of such procedures every five years. During the walkthrough reactor facility, a copy of these operating procedures was not available in the control room. The discussions with the reactor operating personnel showed that there is no procedure for the reactor power operation and shutting down. Start-up check list is available but no step-by-step instructions for conducting facility walkthrough during reactor operation. There are no written operating procedures for the operators' response to all anticipated operational occurrences (for example in case of complete loss of electrical power supply or in case of a stuck of a sample in pneumatic transfer system), and operator
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response to accident conditions such as LOCA and external events need to be revised (see also Section 3.12).
The procedures for adding fuel elements to the reactor core were discussed with the reactor operating personnel. According to the current procedures it is allowed that one fresh fuel element be loaded into the core without prior neutronic analysis. Such a practice may have significant effect on the reactor safety in the view of the possibility to load a fresh fuel element into the core central position, taking into account that the limit on the minimum reactivity shutdown margin is 1$. According to the procedures, the only verification that is required in this case is the measurement of the core excess reactivity. Due to the fact that loading of a fresh fuel element may change the value of the power peak factor and the calibration factor of the nuclear instrumentation, this procedure need to be revised and corrected.
Recommendations:
R10) The operating procedures should be completed to cover all the operations of safety significance including development of procedures for reactor power operation and shutting down, for the facility walkthrough, and for the operators' response to the anticipated operational occurrences and accident conditions. A copy of the operating procedures should be made available at the control room.
R11) The operating procedures for fuel element loading into the core should be revised so as to forbid such an operation unless it is analyzed from safety point of view. Such analysis should verify compliance with the OLCs on the minimum shutdown margin, determine the value of the power peak factor, and evaluate the effect on the calibration of nuclear instrumentation. The newly configured core should not be released for routine operation unless these parameters are evaluated by appropriate measurements.
3.9. MAINTENANCE PERIODIC TESTING AND INSPECTION
The routine maintenance activities are performed by the reactor operators. Some routine maintenance activities and repair are performed by the IJS Technical Service or outsourced. The specialists of the Technical Service are qualified to perform maintenance activities that involve exposure to radiation or contamination. A system for maintenance work permit existsand is subjected to approval of the radiation protection officer.
The reactor manager explained that routine maintenance activities are not performed according to a 'fixed time-table'. Although there are activities of preventive maintenance for some of the reactor SSCs, a comprehensive preventive maintenance programme that covers all the SSCs with the maintenance schedule and type of maintenance activities is not established.
There are maintenance procedures for some activities such as calibration verification of the radiation protection monitors. However, not all the routine maintenance activities are covered by written procedures.
There are deficiencies in the periodic testing (surveillance requirements) programme (see also Section 3.7 on OLCs). Important periodic testing activities (including periodic checks and
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calibration of the measurements channels) are not performed on the basis of written approved procedures.
A large number of in-service inspection activities were performed (or planned) within the programme for the Periodic Safety Review (PSR). The results of the in-service inspection performed on the fuel elements, reactor pool and internal structure showed acceptable physical status of these SSCs.
Recommendation:
R12) A formal preventive maintenance programme should be established and implemented. It should cover all SSCs important to reactor safety. Maintenance procedures, including step-by-step instructions and acceptance criteria should be developed and be referenced in the revised version of the SAR.
Suggestions:
S2) For the effective planning and implementation of the preventive maintenance programme, it is suggested to develop a maintenance schedule that includes all SSCs subject to preventive maintenance, and the planned dates of the relevant maintenance activities.
S3) It is suggested that verification of the SSS be added to the start-up checklist. This refers mainly to the operability check of the SCRAM trips initiated by the nuclear instrumentation (on the reactor power and period).
Good Practice:
GP2) Performing a PSR for the research reactor every 10 years (with the objective and scope similar to the nuclear power plants) is considered as a good practice.
3.10. UTILIZATION AND MODIFICATIONS
There are regulatory requirements for the classification of the proposed experiments and modifications according to their safety significance. The criteria for classification are applied by the operating organization. However, it is not clear according to these procedures if a formal safety analysis should be performed. Experiments are considered as temporary modifications. Although it is not included in written procedures, each experiment is reviewed by the reactor manager and the SC. The current version of the SAR does not have any description on the safety experiments. The OLCs do not include limits on the reactivity worth of experiments.
The practice on establishing a list of forbidden materials to be irradiated including the amount of these materials and their chemical and physical form was discussed with the operating personnel. The reactor personnel indicated that there is no need for such a list as samples to be irradiated are of repetitive nature and are previously analysed. An analysis is performed if a new material is requested to be irradiated.
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Recommendation:
R13) Formal procedures should be established for the planning and implementation of experiments and modifications, which should require that a safety analysis be performed for every proposed modification or new experiments. The OLCs should be revised to include limits on the reactivity worth of experiments (maximum reactivity worth of a single experiments and maximum reactivity worth for all experiments). The revised version of SAR should include a chapter on the safety of experiments.
3.11. RADIATION PROTECTION AND RADIOACTIVE WASTE MANAGEMENT
The operational radiation protection and waste management programme is comprehensive. It covers description of organization and responsibilities of the IJS radiation protection unit, dose limits and constraints, description of the radiation measuring equipment, description of the radiation sources at the reactor facility, classification of areas and zoning, and procedures covering all aspects of operational radiation protection of the TRIGA MARK II reactor.
A radiation protection officer is available at the reactor during the day operation shifts. The radiation protection officer is on-call during the night operation shifts but according to the established rules, should be available as necessary at the reactor building within one hour.
The dose limits and dose constraints are established by the SNSA following the international standards. The personal dosimeters are read monthly. The annual average dose received by an individual from the reactor is less than 1 mSv.
A programme for external radiation monitoring within the reactor facility is established. External dose rate is routinely measured at pre-defined areas within the reactor building. The frequency of measurements is once per week and as necessary according to the performed activities. Contamination levels at pre-defined areas within the reactor building and associated laboratories are also measured. The radiation protection group also measures the radioactivity contents in the primary coolant water, and there is a continuous monitoring of the liquid waste discharged from the reactor. The usual total annual liquid release is less than 10 MBq. Gas effluents are monitored online through stack detectors, and there is a programme for environmental monitoring.
The calibration of the radiation protection measuring equipment is valid and performed according to written procedures. The radiation protection records are complete and there is an annual reporting to the SNSA.
The areas within the reactor building are classified from radiation protection point of view to supervised and controlled areas. The controlled areas are further categorized into two zones. The ventilation system area is classified as supervised area although there is no physical barrier or control of contamination at its borders. During the facility walkthrough, it was observed that the over-coats used in the controlled areas are also used in the supervised areas. No over-shoes are used within the controlled areas.
Systems and procedures are established for collection, segregation and monitoring of radioactive waste generated form the reactor. Limits on waste discharge and clearance levels are established by the national regulations and are complied with.
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Recommendations:
R14) The ventilation system area should be re-classified as a controlled area, or otherwise access without contamination control should be denied. The over-coats used in the controlled areas should not be used in the supervised areas.
R15) Handling of radioactive material during the night operation shifts should not be performed unless the radiation protection officer is available in the reactor building.
Suggestions:
S4) It is suggested to consider establishment of administrative radiation protection rule to use over-shoes in the controlled areas to prevent potential spread of contamination.
S5) It is suggested to develop flooring maps for the reactor building with indication of the area classification from radiation protection point of view. These maps could be included in the revised version of the SAR.
3.12. EMERGENCY PLANNING
The emergency plan of the TRIGA MARK II reactor is part of the IJS emergency plan and was developed in 2011. The reactor emergency plan includes emergency organization and responsibilities, emergency classification system, emergency response, protective measures and training on emergency. Radiological criteria for different emergency classes are established and included in the emergency plan. Protective measures are defined and include sheltering and evacuation of the reactor site. There are written procedures for reactor building evacuation.
The emergency plan considers the PIEs presented in the current version of the SAR. In this regard, it should be noted that these PIEs need to be revised as discussed in Section 3.6 of this report.
Emergency equipment is available at the reactor site and their proper functioning is periodically checked. The equipment includes radiation and contamination detection devices, and decontamination supplies.
Training on emergency is performed frequently. In October 2012, an on-site emergency drill was conducted. The scenario considered was earthquake which led to LOCA and requires evacuation of the reactor site. A process is in place to assess the results of the emergency drills and to incorporate feedback from them into revised emergency plan or procedures. On the basis of the October 2012 emergency drill, a need was identified to establish step-by-step instructions to the operator in case of LOCA. The results of emergency exercises are reviewed by the SC.
Recommendation:
R16) In view of the recommended revision of the safety analysis (see Section 3.6), the emergency plan should be revised accordingly to ensure coherence with the identified PIEs.
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The emergency plan should consider the beyond design basis accident which could be determined using the design basis accident with degradation of the facility (e.g. degradation of the filtration system).
3.13. QUALITY ASSURANCE
The QA programme for the TRIGA MARK II reactor is a part of the QA programme of the IJS. It includes procedures and provisions to control the effective implementation of the quality of the activities related to the reactor operation, utilization and modification.
Suggestion:
S6) It is suggested to establish a document presenting the different disciplines and competencies needed for the operation and utilization of the reactor, the existing qualified human resources to perform the corresponding activities and those to carry out the associated QA control. This will help identify possible gaps for the effective application of QA procedures.
3.14. DECOMMISSIONING
A preliminary decommissioning plan is available for the reactor. Decommissioning of the reactor facility is the responsibility of the IJS. The current version of decommissioning plan include information on decommissioning organization, categorization of generated waste and estimation of its volume, activity and contamination levels.
The reactor management and operating personnel are aware that the decommissioning plan should be developed during the design stage of the reactor and is subjected to periodic revision and amendment through the reactor lifetime. The decommissioning plan is approved by the regulatory body and subjected to periodic review every 5 years.
The operating personnel mentioned that currently there is no national repository site for final disposal of low and intermediate radioactive waste, and there are no acceptance criteria for final disposal of such types of waste. These pose difficulty to develop a more detailed decommissioning plan. Environmental samples are collected and analysed in a yearly basis, and the results of the analysis are submitted to the regulatory body.
Suggestion:
S7) It is suggested to use the records of different environmental radiation measurements to establish a baseline data for the ultimate decommissioning of the reactor facility. In the case of gaps in these data, actions have to be taken to complete them.
3.15. REGULATORY SUPERVISION
The regulatory supervision of the nuclear installations in Slovenia (Krsko NPP and TRIGA MARK II reactor) is ensured by an independent regulatory body (SNSA). For the TRIGAMARK II reactor there are appropriate regulations covering specific requirements for research
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reactors, application of the graded approach, licensing process, inspection and reporting, design and operation, environment monitoring, and radioactive waste.
The duration of the operating license validity for the TRIGA Mark II reactor is limited to ten years. Its renewal by the SNSA is based on the results of a PSR. The programme for the first PSR was approved in 2011 and is currently on-going for the reactor.
The SNSA performs regular inspections of the reactor according to an inspection plan established each year. The frequency of inspections is at minimum once per year.
In 2011 and following the accidental fire in one hot cell in the controlled area of the reactor, the SNSA conducted three inspections with a focus on QA, SAR and emergency preparedness. The results of the inspections are formalized in a report and communicated to the operator for follow up actions.
In summary, the regulatory supervision of the TRIGA reactor is effective and based on a set of comprehensive regulations applied with the use of the graded approach.
Suggestion:
S8) It is suggested to include in the topics of future inspections the verification of application of OLCs, in particular those related to periodic testing.
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ANNEX 1: AGENDA
Sunday - 11 November 2012 Team meeting (Hotel): Discussions of the general comments on the received documents
Monday 12 November 201209:00-12:00 – Introduction (Counterparts, IAEA).
– Presentation on the current safety and utilization status of the TRIGA reactor(Counterparts).
– General comments on the information package received from the Counterparts– Walkthrough at the reactor facility (All)
12:00-13:00 Lunch break
13:00-17:00 – Continuation of the walkthrough – Operating organization and reactor management– Safety committeeReview and discussion with the Counterparts
Tuesday – 13 November 2012
09:00-09:30 – Briefing to the Counterparts on the finding of the previous working day (A. Shokr and Main Counterpart)
09:00-12:00 – Training and qualification of personnel– Safety Analysis Report and safety analysis – Operational limits and conditions
Review and discussion with the Counterparts
12:00-13:00 Lunch break
13:00-17:00 – Conduct of operations including operating procedures, records and reports– Maintenance, periodic testing and inspection programme
Review and discussion with the Counterparts
Wednesday – 14 November 2012
09:00-09:30 – Briefing to the Counterparts on the finding of the previous working day (A. Shokr and Main Counterpart)
09:00-12:00 – Radiation protection and waste management programme
– Emergency planning
Review and discussion with the Counterparts
12:00-13:00 – Lunch break
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13:00-17:00 – Quality assurance programme
– Decommissioning plan
Review and discussion with the Counterparts
Safety of experiments and modifications
Review and discussion with the Counterparts
Thursday – 15 November 2012
09:00-10:30 – Discussions on emergency planning (Continuation: IAEA and Counterparts)
10:30- 12:30 – Review and discussion with the Counterparts: Regulatory supervision (IAEA and Counterparts from SANS)
12:30-13:30 – Lunch break
13:30-17:00 – Drafting of the mission report (IAEA mission team)
Friday – 16 November 2012
09:00-12:00 – Finalization of the draft mission report, and discussions with the Counterparts
12:00 – Exit Meeting: Conclusions and recommendations of the mission.
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ANNEX 2: LIST OF PARTICIPANTS
IJS and SNAS
Name Organization and function E-mail address
Igor Grlicarev SNSA [email protected]
Andreja Persic SNSA, Nuclear Safety Division
Tomaz Nemec SNSA, Nuclear Safety Division
Ljubo Fabjan IJS, Head of QA Unit [email protected] Gyergyek IJS, RIC [email protected] Stepisnik IJS, RIC [email protected]
Luka Snoj IJS, Head of TRIGA MARK II Reactor
Borut Smodis IJS, Head of the RIC [email protected]
Tinkara Bucar Radiation Protection Unit, IJS
Ales Cesar Technical Service Unit, IJSJose Hatko Security Coordinator, IJSRadivoje Sucur Safety at Work Unit, IJS
Review team
Name Organization and function E-mail addressA. M. Shokr IAEA, SH-RSS/NSNI/IAEA [email protected]
H. Abou Yehia IAEA, RRSS/NSNI/IAEA [email protected]
H. Hirshfeld IRR-1, Soreq Nuclear Center, Israel
V. Garea INVAP, Head Dept. of Safety and Environmental Impact
A. Jraut CNESTEN, Safety Analysis Unit
