BARC/2018/E/005 BARC/2018/E/005
TRANSITION FROM OPERATION TO DECOMMISSIONING OF CIRUS RESEARCH REACTOR by Rakesh Ranjan, S. Bhattacharya, C.G . Karhadkar, Prasit Mandal, M.K. Ojha Reactor Operations Division
2018 BARC/2018/E/005
GOVERNMENT OF INDIA DEPARTMENT OF ATOMIC ENERGY BARC/2018/E/005
TRANSITION FROM OPERATION TO DECOMMISSIONING OF CIRUS RESEARCH REACTOR by Rakesh Ranjan*, S. Bhattacharya, C.G . Karhadkar, Prasit Mandal, M.K. Ojha [email protected]* Reactor Operations Division Reactor Group
BHABHA ATOMIC RESEARCH CENTRE MUMBAI, INDIA 2018 BARC/2018/E/005
BIBLIOGRAPHIC DESCRIPTION SHEET FOR TECHNICAL REPORT (as per IS : 9400 - 1980)
01 Security classification : Unclassified
02 Distribution : External
03 Report status : New
04 Series : BARC External
05 Report type : Technical Report
06 Report No. : BARC/2018/E/005
07 Part No. or Volume No. :
08 Contract No. :
10 Title and subtitle : Transition from operation to decommissioning of Cirus research reactor
11 Collation : 82 p., 39 figs., 9 tabs.
13 Project No. :
20 Personal author(s) : Rakesh Ranjan; S. Bhattacharya, C.G. Karhadkar; Prasit Mandal; M.K. Ojha
21 Affiliation of author(s) : Reactor Operations Division, Bhabha Atomic Research Centre, Mumbai
22 Corporate author(s): Bhabha Atomic Research Centre, Mumbai - 400 085
23 Originating unit : Reactor Operations Division, Bhabha Atomic Research Centre, Mumbai
24 Sponsor(s) Name : Department of Atomic Energy
Type : Government
Contd... BARC/2018/E/005
30 Date of submission : April 2018
31 Publication/Issue date : April 2018
40 Publisher/Distributor : Head, Scientific Information Resource Division, Bhabha Atomic Research Centre, Mumbai
42 Form of distribution : Hard copy
50 Language of text : English
51 Language of summary : English
52 No. of references : 23 refs.
53 Gives data on : Abstract : Cirus, a 40 MW (Thermal), tank type reactor had been in operation since 1960. The 60 reactor was shut down from 1997-2003 for refurbishment based on a comprehensive ageing assessment. It was restarted in year 2003 with significant increase in availability and safety margins and was shut down permanently on 31st December, 2010. Prior to shutting down the reactor, a core management programme was implemented to maximize fuel utilization. After shutting down the reactor, the core was unloaded of all spent fuel assemblies, isotope production assemblies, experimental assemblies and shut off rods. Subsequently the entire spent fuel was sent for reprocessing. Various reactor systems were maintained initially in wet preservation mode and chemistry of process fluids was monitored and maintained. Subsequently the systems were brought to dry preservation mode after draining of process fluids and drying the systems. This state has reduced surveillance requirements as well as cost in maintaining them. Only need based operation of auxiliary systems is being done at present. Radiological and industrial hazards were minimized by taking suitable measures. Suitable modifications in systems were carried out to reduce surveillance. With this, manning of reactor in round-the-clock shifts has been stopped since 10th July 2017 after approval of regulator. Reactor is manned on normal working days in general shift hours. For Cirus reactor, deferred dismantling (decommissioning) strategy is considered as the best option. Near term, mid-term and long-term activities have been identified as a part of deferred decommissioning programme. Detailed radiation mapping to assess radiation field inside core has been carried out. Preliminary waste volume estimation and characterization of radionuclides has been carried out. Detailed radiological characterization has been started. Access control to various areas in and around the reactor building is being maintained in the same manner as it was during the operating days of the reactor. All the areas of potentially high radiation field and contamination are under access control with lock and key. A two-tier set-up of experts having experience in operation and maintenance of reactor, waste management and radiological protection has been made to prepare documents for decommissioning of the reactor. All plans for decommissioning are subjected to regulatory review. Decommissioning Superintendent co-ordinates and executes the plans
70 Keywords/Descriptors : CIRUS REACTOR; REACTOR SHUTDOWN; REACTOR DECOMMISSIONING; MAPPING; RADIOACTIVE WASTE MANAGEMENT; REACTOR SAFETY; RADIOACTIVE WASTE STORAGE 71 INIS Subject Category.: S21
99 Supplementary elements : TRANSITION FROM OPERATION TO DECOMMISSIONING OF CIRUS RESEARCH REACTOR
Rakesh Ranjan*, S. Bhattacharya, C.G. Karhadkar, Prasit Mandal, M.K. Ojha Reactor Operations Division, Reactor Group Bhabha Atomic Research Center, Mumbai, India [email protected]*
Abstract
Cirus, a 40 MW (Thermal), tank type reactor utilizing heavy water as moderator, graphite as reflector, demineralized light water as primary coolant and natural uranium metal as fuel; had been in operation since 1960. After about three decades of operation, the reactor exhibited signs of ageing. The reactor was shut down from 1997-2003 for refurbishment based on a comprehensive ageing assessment. It was restarted in year 2003 with significant increase in availability and safety margins and was shut down permanently on 31st December, 2010. After shutting down Cirus reactor permanently, the core was unloaded of all spent fuel assemblies, isotope production assemblies, experimental assemblies and shut off rods. Subsequently the entire spent fuel was sent for reprocessing. Radio-isotopes were delivered to Radio Pharmaceutical Division for processing and utilization. Irradiation assemblies were sent to Waste Management Division for storage and disposal. At present, Cirus site doesn’t have any irradiated fuel assembly under storage, except a few irradiated test fuel assemblies of Pressurized Water Loop (PWL), stored in water bays of Rod Cutting Building (RCB). Detailed radiation mapping to assess radiation field inside core has been carried out. Preliminary waste volume estimation and characterization of radionuclides has been carried out. Detailed radiological characterization has been started. Various reactor systems were maintained initially in wet preservation mode and chemistry of process fluids was monitored and maintained. Subsequently the systems were brought to dry preservation mode after draining of process fluids and drying the systems. This state has reduced surveillance requirements as well as cost in maintaining them. Only need based operation of auxiliary systems is being done at present. Radiological and industrial hazards were minimized by taking suitable measures. Suitable modifications in systems were carried out to reduce surveillance. With this, manning of Cirus in round-the-clock shifts has been stopped since 10th July 2017 after approval of regulator. Reactor is manned on normal working days in general shift hours. A Supervisor assisted by 2-3 technicians is in charge of the surveillance activities and execution of other planned jobs. A common Radiological Hazard Control (RHC) unit for Dhruva and Cirus facilities provides radiological safety / health physics coverage. A two-tier set-up of experts having experience in operation and maintenance of reactor, waste management and radiological protection has been made to prepare documents for decommissioning of the reactor. The lower tier headed by Decommissioning Superintendent is entrusted with preparation of the documents. The upper tier headed by Director / Associate Director, Reactor Group reviews and approves the documents. All plans for decommissioning are subjected to regulatory review. Decommissioning Superintendent co-ordinates and executes the plans. Access control to various areas in and around the reactor building is being maintained in the same manner as it was during the operating days of the reactor. All the areas of potentially high radiation field and contamination are under access control with lock & key. For Cirus reactor, deferred dismantling (decommissioning) strategy is considered as the best option. Near term, mid-term and long-term activities have been identified as a part of deferred decommissioning programme. Dismantling of reactor structure and core components is envisaged after about 30-35 years when the dominant radionuclide Co-60 will decay significantly for ease of dismantling and handling of radioactive components. Peripheral systems and components will be dismantled and disposed off in the initial years. Ventilation system and other auxiliary systems required for dismantling operations, radiological and industrial safety, security of reactor, etc. has been kept operational. Optimum surveillance is being maintained. Resources available in the nearby Dhruva reactor, which can contribute in long term surveillance requirements, help in deciding deferred decommissioning as a preferred strategy. With implementation of suitable steps, the reactor has been brought to a ‘Safe Storage’ state. Short term activities which are planned to be executed in the next few years have been identified which include isolation and truncation of systems to reduce surveillance requirements, disposal of inactive components, disposal of low contaminated equipment / components to waste management facility, etc. To gain experience with decommissioning, dismantling of some inactive systems and low active systems has been started. The space created from this activity will be used for installation of new experimental facilities. In Cirus reactor hall, several new facilities have been installed to utilize the space. A new Technical Specifications for ‘Safe Storage’ state of the reactor has been prepared. Documentation for approval of decommissioning plan is under preparation for submission to regulatory authorities. CONTENT
1 INTRODUCTION ……………………………………………………………………… 1 2 BRIEF DESCRIPTION OF CIRUS REACTOR………………………………………….. 2 3 FACTORS AFFECTING DECOMMISSIONING DECISION AT CIRUS……… 5 4 MAJOR CHANGES DURING THE TRANSITION PERIOD ………...… 6 5 ORGAIZATION AND MANAGEMENT DURING THE TRANSITION PERIOD 7 6 IMPLEMENTATION ISSUES……..……………………………………………………... 10 6.1 Characterization and Inventory of Radioactive and Hazardous Materials…………………. 10 6.2 Radiation Mapping of Core Components ………………………………………………. 11 6.3 Consideration for Functional Systems……………………………………………………... 15 7 THE IMPACT OF THE TRANSITION AND DECOMMISSIONING ON HUMAN RESOURCES …………………………………………………………………………. 15 8 COST REDUCTION BY RETIREMENT OF SYSTEMS OR RECONFIGURATION 18 9 DEVELOPMENT OF TECHNIQUES AND TOOLS…………………………………… 18 10 WASTE MANAGEMENT …………………………………………………………… 20 11 MANAGEMENT/ADMINISTRATIVE ISSUES………………………………………….. 29 11.1 Taking a Comprehensive Inventory of Radioactive and Hazardous Materials ………. 29 11.2 Review of Purchasing Policy and Spares………………………………………………… 29 11.3 Records to Support Decommissioning ………………………………………………. 30 11.4 Interaction with Stakeholders…………………………………………………………… 32 11.5 Training to Support the Transition …………………………………………………… 33 12 IMPLEMENTATION OF TRANSITION ACTVITIES………………………………… 34 12.1 Core Management and Removal of Spent Fuel………………………………………….. 35 12.2 System Cleanout Operation………………………………………………………………... 36 12.2.1 Removal of Moderator and Cover Gas…………………………………………………………… 36 12.2.2 Draining of Primary Coolant……………………………………………………………………… 37 12.2.3 Draining of Pressurized Water Loop (PWL) System…………………………………………. 40 12.2.4 Thermal Shields Recirculation System (TSRS)………………………………………………... 40 12.3 Ventilation System………………………………………………………………………… 41 12.4 Sumps and Liquid Waste Transfer System……………………………………….……….. 41 12.5 Service Water & Spray Pond / Machinery Cooling Water (MCW) System...... ……. 42 12.6 Mechanical Handling Equipment………………………………………………………….. 43 12.7 Rod Cutting Building (RCB)………………………………………………………………. 43 12.8 Power Supply System…………………………………………………………………….. 43 12.9 Control & Instrumentation Systems……………………………………………………... 45 12.10 Seawater System………………………………………………………………………….. 45 12.11 Compressed Air System…………………………………………………………………. 45 12.12 30 T /day Low Temperature Evaporation Desalination Plant……………………………... 46 12.13 Reduction of Hazards……………………………………………………………………… 46 12.14 Reduction in Surveillance and Stoppage of Round-the Clock Manning of the Reactor...… 47 13 SAFETY CONSIDERATIONS DURING TRANSITION FROM OPERATION TO DECOMMISSIONING ……………………………………………………………… 48 13.1 Fuel Handling and Storage………………………………………………………………… 48 13.2 Drainage of Systems………………………………………………………………………. 48 13.3 Cleaning and Decontamination……………………………………………………………. 49 13.4 Conditioning and Removal of Operational Waste………………………………………. 50 13.5 Retirement, Reconfiguration and Planning for the Provision of New Systems……..……. 50 13.6 Changes to Confinement Barriers…………………………………………………………. 50 13.7 Safety Analysis…………………………………………………………………………….. 51 13.8 Exposure of Personnel……………………………………………………………………. 51 13.9 Considerations for Long Term Storage……………………………………………………. 52 13.10 Other Accidents Possible during the Transition Period…………………………………… 52 14 TREATMENT, CONDITIONING, STORAGE AND/OR DISPOSAL OF WASTE DURING THE TRANSITION PERIOD…………………………………………………. 52 15 DECONTAMINATION OR FIXING OF CONTAMINATION………………………… 54 16 CHARACTERIZATION AND INVENTORY OF RADIOACTIVE AND HAZARDOUS MATERIALS…………………………………………………………. 56 17 PREPARATION OF A FACILITY’S ROOMS AND BUILDINGS DURING TRANSITION…………………………………………………………………………… 60 18 PROTECTION FROM EXTERNAL OR INTERNAL EVENTS…….....……………….. 61 19 REMOVAL OF MINOR COMPONENTS……..…………………………………………. 61 20 COST OF TRANSITION ACTIVITIES…….…………………………………… 62 21 UTILIZATION OF SPACE DURING SAFE ENCLOSURE…...……………………… 64 22 CONCLUSIONS……………………………………………………………………….. 65 ACKNOWLEDGEMENTS………………………………………………………………… 67 REFERENCES…………………………………………………………………………………….. 68 List of Figures Sr. No. Title Page
Fig.1 Decommissioning related activities during the life cycle of a nuclear reactor 2
Fig.2 Cirus Reactor Structure Cross-section 4
Fig.3 Simplified flow circuits of Cirus Reactor 4
Fig.4 Cirus staff in decorated control room on the last day of its operation 5
Fig.5 Typical example of a functional organization during the transition period 8
Fig.6 Decommissioning organization structure as per AERB 9
Fig.7 Decommissioning management structure at Cirus 10
Fig.8 Schematic of SS tubes in upper tube sheet of Reactor vessel 12
Fig.9 (a) Staffing trend during transition, SE and dismantling; (b) during the 16 transition to immediate dismantling
Fig.10 Operations staff trending at Cirus during transition from operations to 17 decommissioning
Fig.11 Schematic arrangement of Cutting Tool 19
Fig.12 Sample Cut from J-09 Tube of RV 19
Fig.13 East Thermal Column 20
Fig.14 Hole after removing the plugs 20
Fig.15 Solid radioactive waste drum scanning system. 28
Fig.16 Simplified schematic diagram of moderator and cover gas system 36
Fig.17 Tritium activity (Bq/M3) in exhaust air from reactor. 37
Fig.18 Primary coolant system schematic 39
Fig.19 DM water purification plant 40
Fig.20 Ultrasonic decontamination machine 40
Fig.21 PWL cooling water line terminated in cable room 42 Fig.22 SS containers for storing U-slurry under water 43
Fig.23 Schematic of U-slurry collection set-up 43
Fig.24 Electricity consumption after permanent shut down 44
Fig.25 Isolated trip and alarm panel 45
Fig.26 Relevant alarms have been put in one panel 45
Fig.27 Desalination Plant 46
Fig.28 Area after dismantling of Desalination plant 46
Fig.29 Dismantling and disposal of acid slurry 47
Fig.30 Space after disposal of acid and alkali tanks 47
Fig.31 Experimental facilities in reactor hall 53
Fig.32 Shielding sections of spent fuel rods 54
Fig.33 Horizontal flask dismantling 54
Fig.34 Cover gas purification system 62
Fig.35 Area after dismantling Cover gas purification system 62
Fig.36 I-131 processing facility at Radio- pharmaceutical Division. A similar facility 65 under installation at Cirus
Fig.37 Mock-up fuel fabrication facility of AHWR 65
Fig.38 PCF, under dismantling 65
Fig.39 PHM Lab. being set-up in PCF room 65 List of Tables
Sr. No. Title Page
Table.1 Characteristics of Cirus Reactor 3
Table.2 Comparison of the Operating and Decommissioning Regimes 7
Table.3 Radiation Dose Rates (mGy/h) in Selected Fuel Rod Positions (year 2011) 13
Table.4 Estimate of solid waste excluding reactor core structure 21
Table.5 Prominent Isotopes of concern in Cirus research reactor decommissioning 27
Table.6 Documents and their Storage Location 32
Table.7 Data for decontamination of fuel channel isolating valves 56
Table.8 Typical characterization data of different components 58
Table.9 Major Radio-nuclides in Various Systems 59 1 INTRODUCTION [1]
The transition period from facility operation to implementation of a decommissioning strategy is an important one. During this period a number of plans and modifications are made to adapt a facility to new objectives and requirements. Transition activities take place between operation and placement of the facility in a safe and stable condition preparatory to safe enclosure (storage) and/or dismantling. Typically, these activities include defueling of reactors, retirement of equipment and systems, radiological and waste characterization, operational waste treatment and removal of minor components. Generally, removal or dismantling of major components during Safe Enclosure (SE) are excluded. However, activities carried out during the transition period depend upon the type of facility and the regulatory regime. The objective of the transition period is to plan and implement these activities in a timely manner. A cultural change is also needed to reflect different management and working practices. It is essential that planning for the transition and decommissioning begin during operation and that activities be implemented as soon as possible after permanent shutdown to ensure a controlled transition and the best use of resources. A key to the success of the transition period is the training and preparation of facility personnel. This includes, in particular, utilizing operating staff whose knowledge of the facility and its systems is invaluable during this transition period. In addition, a number of strategic and administrative issues are addressed before or immediately after permanent shutdown of the plant to support planning for decommissioning and to reduce the burden of operational requirements.
Typical transition activities are [2]: (i) Sale, further use, recycling or dismantling of usable fissile/fertile materials. (ii) Removal of spent fuel and other fissile/fertile material from the plant. (iii) Removal of spent fuel and other fissile/fertile material from the site (if applicable). (iv) Stabilization, treatment and/or removal of potentially unstable materials or wastes. (v) Reduction or elimination of the potential for fire or explosions from violent chemical reactions or nuclear criticality. (vi) Completion of cleanout operations of systems, lines and other equipment not needed in the future that have the potential for significant radioactive and chemical material inventory. (vii) Neutralization and disposal of hazardous chemicals and oil in storage. (viii) Review, using the safety assessment, of changes in the configuration and status of systems and structures as a result of transition activities, e.g. reducing redundancies in systems and structures. (ix) Revision of operating requirements and controls as appropriate to changed conditions; this should also include the number of personnel required to maintain the appropriate safety standards. (x) Installation and/or verification of sufficient barriers to prevent the spread of contamination. (xi) Verification of appropriate safeguards and security. (xii) Checking and updating of relevant facility drawings and other documents to reflect changes that have been made during the operational period and/or the transition period.
1 (xiii) Training and awareness of facility staff for their future work and roles.
(Fig.1) shows the schematic of typical decommissioning related activities during the life cycle of a nuclear reactor.
Fig.1. Decommissioning related activities during the life cycle of a nuclear reactor [1]
The activities carried out during the transition period have two main objectives: (i) The efficient operational conversion of the facility from its original mission to one in which operations, surveillance and maintenance are reduced, consistent with the lower safety risk, the systematic reduction in hazard and the need to cost effectively prepare for either SE or immediate dismantling. (ii) The preparation of a detailed decommissioning plan, which requires the most current information available regarding the condition of systems, structures, components and materials. For example, a full radiological characterization is required to provide data for updating the decommissioning plan and the waste management strategy.
Transition activities serve to achieve a safer and more economical configuration, e.g. by reducing the inventory of radiological material, dealing with spent fuel, removing hazardous chemicals, focusing surveillance and maintenance on plant features needed to control contamination and other hazards.
2 BRIEF DESCRIPTION OF CIRUS REACTOR Cirus, a 40 MWt, tank type research reactor utilizing heavy water as moderator, graphite as reflector, demineralized light water as primary coolant and natural uranium metal as fuel; had been in operation since 1960. Sea water is used as secondary coolant. (Table.1) shows characteristics of Cirus reactor while (Fig.2) shows Cirus Reactor Structure Cross-section. The reactor block is located inside a steel containment and it houses the reactor vessel, graphite
2 reflector, cast iron thermal shields, aluminum and steel thermal shields and concrete biological shields. Reactor Vessel (RV) is a cylindrical Aluminium tank of 267 cm diameter and 320 cm height. It has 3” thick cylindrical disk shaped top and bottom tube sheets. 199 vertical tubes of different diameter, arranged in a hexagonal lattice, are expanded and rolled in the tube sheets and pass through reactor vessel to permit insertion of in-pile assemblies. The tube sheets are cooled by water flowing through holes drilled along chords. Water inlet to the holes is through SS tubes connected to primary coolant pipe. (Fig.3) shows simplified schematic of major systems of Cirus reactor.
Table.1. Characteristics of Cirus Reactor
Reactor type Tank Thermal power (kW) 40,000 First criticality 10thJuly 1960 Reactor shut down for refurbishment Sept. 1997 Reactor start-up after refurbishment Oct. 2003 Permanent shut down 31stDecember 2010 Maximum Neutron Flux 6.6x1013 n/cm2/sec Fuel Natural Uranium Coolant Light water Moderator Heavy water Reflector Graphite
Shut down device B4C rods
Since the late 80’s, Cirus started showing signs of ageing necessitating extra efforts for maintenance and the reactor availability factor started declining gradually. Detailed ageing studies were undertaken during 1990-95 and SSCs which required replacement / upgradation for trouble free operation of reactor for next 10-15 years were identified. Reactor was shut down in 1997 for refurbishment work and re-started in 2003. Post refurbishment, reactor operation was excellent with an availability factor of > 80%. Reactor was shut down permanently on 31st December 2010. Subsequently core was unloaded and preparatory works for decommissioning were initiated. (Fig.4) is a snapshot of decorated control room on the last day of its operation.
3 Fig.2. Cirus Reactor Structure Cross-section
Fission heat was removed by primary coolant provided by four pumps operating in parallel. Primary coolant in turn was cooled by sea water. When the primary coolant pumps were not operating, decay heat was removed by DM water provided from a concrete tank (called Ball Tank due to its shape) under gravity. Helium was used as cover gas above heavy water moderator. Sea water was used to remove heat from moderator. Primary coolant pipe lines were primarily made of carbon steel while that of moderator and cover gas system were of stainless steel.
Fig.3. Simplified flow circuits of Cirus Reactor
4 Fig.4. Cirus staff in decorated control room on the last day of its operation
3 FACTORS AFFECTING DECOMMISSIONING DECISION AT CIRUS Cirus reactor being located at BARC is benefitted from the very wide knowledge base of BARC. Expertise in remote handling technologies, decommissioning techniques and waste and material management technologies is available. However not much experience in decommissioning of a major nuclear facility is available. In that sense, decommissioning of Cirus reactor may be treated as ‘first of a kind’ project.
A major issue which needs to be addressed is that of decommissioning regulations. Atomic Energy Regulatory Board (AERB) regulates nuclear facilities other than BARC. BARC has its own regulatory framework headed by BARC Safety Council. Decommissioning regulations are in nascent stage. Currently regulation of decommissioning activities is being carried out under the framework for normal operation of nuclear installations. Adequate regulations for the clearance of sites or materials exist.
The manpower has been successfully reoriented for decommissioning operations. The rich knowledge and experience of the operating staff familiar with various reactor systems was very handy in bringing the systems to a safe storage status. After achieving that status, most of the
5 younger manpower was assigned to new facilities for operation and maintenance. Experienced manpower retained at Cirus has been reoriented to work in project mode to suite decommissioning requirements. None of the manpower was retrenched. However loss of manpower due to superannuation in future is an issue which needs to be addressed. Bringing back the experienced manpower from other facilities is an option or fresh manpower can be employed in which case they would need to be trained properly.
4 MAJOR CHANGES DURING THE TRANSITION PERIOD [1], [3] During the transition, as the experienced plant personnel deplete either due to superannuation or transfer to other plant or leaving the plant for better avenues, use of contract workers to make up for any shortfalls resulting from the loss of experienced staff is normal. Even otherwise, specialized contractors are needed to carry out some jobs which are not of normal nature carried out during operational phase of the reactor. However, it is vital that the plant management retain enough suitably qualified and experienced personnel to understand and work toward the plant’s safety, and to be an ‘intelligent customer’ of these contractors. This is especially important during the transition period if the numbers of permanent staff are declining. Many historical aspects of plant design and operation which need to be accessed during the transition period are known only to individuals and may not have been recorded in documents. These people are therefore important during the transition period when their knowledge and experience is required. (Table.2) compares the characteristics during normal operation and decommissioning regimes.
6 Table.2. Comparison of the Operating and Decommissioning Regimes [1], [4]
Operations Decommissioning Reliance on permanent structures for the Introduction of temporary structures to assist operating life of the facility dismantling Safety management systems based on an Safety management systems based on operating nuclear facility decommissioning tasks Production oriented management objectives Project completion oriented management (except perhaps in research facilities) objectives Routine training and refresher training Retraining of staff for new activities and skills or use of specialized contractors Permanent employment with routine objectives Visible end of employment — refocus of the staff’s work objectives Established and developed operating Change of regulatory focus regulations Predominant nuclear and radiological risk Reduction of nuclear risk, changed nature of radiological risk, significantly increased industrial risk Focus on functioning of systems Focus on management of material and radioactivity inventory (e.g. for waste minimization) Repetitive activities One-off activities Working environment well known Working environment unknowns possible Routine lines of communication New lines of communication Low radiation/contamination levels relatively Low radiation/contamination levels important unimportant for material clearance Access to high radiation/contamination areas Access to high radiation/contamination areas unlikely or for a short time for extended periods Routine amounts of material shipped off-site Larger amounts of materials shipped off-site Relatively stable isotopic composition Isotopic composition changing with time
5 ORGAIZATION AND MANAGEMENT DURING THE TRANSITION PERIOD The decommissioning of a large nuclear facility with the activities involved in the transition is a major project. The best project management practices, tools and techniques, as well as quality assurance processes, are vital. During the transition period many operational hazards are removed in preparation for SE and/or immediate dismantling. This includes removal of the spent fuel, draining of systems, Post-Operational Cleanout (POCO) and removal of waste generated during operation. The management structure at all times reflects the circumstances and continuing responsibility of the licensee for the licensed site. At the start of the transition period, the organization inevitably remains that of the operational phase. Even in cases where a new
7 operating organization takes over for decommissioning, it is likely that most of the operating staff is retained and their roles change to reflect the activities during the transition period, as depicted in (Table.2). (Fig.5) shows a typical functional organization as it might be modified for transition and decommissioning projects and tasks. In addition to the facility personnel, contractors will be assigned to some jobs, particularly during dismantling. Owner/ licensee
Site manager Management of change strategy
Plant Operational Engineering Transition/ Operations services services decommissioning projects
Operation Personnel System shutdown Implementation of Maintenance Procurement Plant change management Health physics & Security reconfiguration Planning chemistry Information New equipment Co-ordination Quality assurance Administration Quality Assurance Control Quality Environmental impact Assurance assessment Quality Assurance
Transition/ decommissioning implementation
Fig.5. Typical example of a functional organization during the transition period [1]
(Fig.6) depicts a typical decommissioning organization as per AERB [5]
8
Fig.6. Decommissioning organization structure as per AERB [5]
Decommissioning Management Structure at Cirus: A two tier set-up of experts having experience in operation and maintenance of reactor, waste management and radiological protection has been made to prepare documents for decommissioning of the reactor. The lower tier headed by Decommissioning Superintendent is entrusted with preparation of the documents and the upper tier headed by Director / Associate Director, Reactor Group reviews and approves the documents. All plans for decommissioning are subjected to regulatory review. The Decommissioning Superintendent co-ordinates and executes the plans. (Fig.7) shows the decommissioning management structure adopted at Cirus.
9
Fig.7. Decommissioning management structure at Cirus
6 IMPLEMENTATION ISSUES
6.1 Characterization and Inventory of Radioactive and Hazardous Materials The characterization and establishment of an inventory of radioactive and hazardous materials within the facility involves surveys of existing data, calculations, in situ measurements and/or sampling and analysis. A database is established to provide significant input into the decommissioning planning process and the development of successful implementation plans. With this database, assessment of the following options and consequent decisions are taken: (a) Operating techniques: decontamination processes, dismantling procedures (hands-on, semi-remote or fully remote) and the required equipment; (b) Radiological and industrial protection of the workers, the public and the environment; (c) Waste management, waste classification and disposal options; (d) Discharge authorization; (e) Cost profiles. At the beginning of the transition period, adequate information was collected to assess the radiological status of the facility and the nature and extent of any other hazardous materials present. Data collected during this initial characterization period is based on information available at the time of final shutdown, including historical operating records. A rough estimate about the contaminated land was also made. Samples from activated materials such as Graphite (reflector), carbon steel (coolant piping and heat exchangers), Stainless Steel (moderator and cover gas), etc. has been collected and being analysed in laboratories. A dedicated
10 radio-characterization laboratory is planned to be established to generate extensive databank about the physical, chemical and radiological conditions of the facility, including contaminated land. This information will serve as the technical basis for work and project decisions, including cost estimates, exposure estimates, risk evaluation, waste management, scheduling and workforce requirements, particularly with respect to radiological exposures. Since characterization requires time, money and dose commitment, it should be optimized to meet the above objectives. As this information should be updated on a regular basis, it is important that the database remain active during the entire decommissioning period [6].
6.2 Radiation Mapping of Core Components [7]
Estimation of radiation dose rate levels in the reactor core structure of Cirus reactor after ten months of permanent shut down of the reactor was carried out. Inaccessible locations in the reactor structure were monitored for radiation dose rate using extendable cable type GM based high range radiation monitor and a diode based high range radiation monitor. Radiation dose rates were measured at 21 specified elevations of reactor structure by putting radiation probe through empty reactor vessel tubes. These measurements were carried after removal of all irradiated fuel rods, shut off rods, tray rods and other assemblies in various core positions. Higher radiation dose rates were observed in the 5” gap between lower Al thermal shield and lower tube sheet, at lower tube sheet, upper tube sheet and in 7” gap above upper tube sheet. Radiation dose rate in the 5” gap below lower tube sheet varied between 3.0 - 31 Gy/h. Radiation dose rate in the 7” gap above upper tube sheet varied between 4.5 - 25 Gy/h. Radiation dose rates in the reactor vessel region varied between 1.25 – 3.0 Gy/h. High radiation dose rates just above and below the reactor vessel are mainly due to presence of activated metallic components such as bunch of SS tubes running near above positions (Fig.8). These high radiation dose rates are mainly due the presence of Co-60 in SS tubes. Radiation dose rates due to loose contamination were insignificant as compared to activation of structural components. During radiation dose rate mapping of horizontal experimental beam holes, it was observed that radiation dose rate increased radially from outer face of outer graphite reflector to the reactor vessel side. In all experimental beam hole positions radiation dose rate near reactor vessel was observed to be 0.7 - 0.9 Gy/h. The maximum radiation dose rates observed between outer graphite reflector and inner cast iron thermal shield is due to activation of cast iron thermal shield. (Table.3) shows radiation field inside some selected core positions. In view of the presence of such high radiation dose rates above and below the reactor vessel region, it is prudent that reactor core structure is dismantled only after significant reduction in radiation dose rates. However in case of requirement of immediate dismantling/decommissioning, necessary remote handling tools will be deployed for carrying out above activities with minimum radiation exposure conditions.
11
Fig.8. Schematic of SS tubes in upper tube sheet of Reactor vessel
Core status during radiation field measurement: status of core during radiation mapping was as follows:
Following assemblies were not in core: ‐ All fuel assemblies ‐ All isotope tray rods including adjuster rods and their outer sheaths ‐ Irradiation test assemblies in PWL ‐ All J-rod assemblies i.e. Thoria / Thorium / Co-slug rods ‐ Pneumatic carrier rod from J-rod annulus ‐ All shut-off rod assemblies including barrels
Following assemblies were inside the core: ‐ Activated Zircolloy-2 test section in PWL (N-11) position ‐ Aluminium dummy assemblies in 29 positions; full of DM water ‐ Activated upper and lower sleeves (Al) in 5 nos 4” positions
12 ‐ Activated upper and lower plugs (Al) in Central Thimble (J-15) position ‐ Activated upper and lower plugs (Al) in plugged positions (H-28 and L-09) The reactor vessel was completely free of moderator. Coolant flow through dummy assemblies @200 igpm was continuously maintained. Upper and lower tube sheets and steel and Aluminium thermal shields were full of water.
Under similar conditions, in year 1997 radiation mapping of core had been carried out. However the maximum range of the Teletector used at that time was 1000 R/hr. Cirus refurbishment outage had been extensively used for characterization with a view for planning of its decommissioning in future. That data along with the data under generation during the post operational period after year 2010 would serve as the technical basis for work and project decisions, including cost estimates, exposure estimates, risk evaluation, waste management, scheduling and workforce requirements, particularly with respect to radiological exposures.
Table.3. Radiation Dose Rates (mGy/h) in Selected Fuel Rod Positions (year 2011)
Figures in bracket are of year 1997.
Radiation Dose Rate (mGy/h) Position Location C-07 Q-25 R-22 No. D-22 E-15 F-08 B-8 (C-05)
Revolving 003 0.03 0 0.03 0.03 0.03 floor 0.03 0.03
Upper Header 0.05 0.03 0.05 0.05 0.06 0.07 1 Room (0.02) (0.03) (0.05) 0.07
Upper Header 0.05 0.05 0.07 0.1 0.1 2 Room (0.05) (0.07) (0.07) 0.1 0.05
Biological 0.07 0.05 0.07 0.1 0.1 3 shield (0.01) (0.01) (0.02) 0.15 0.07
Biological 0.1 0.07 0.1 0.15 0.2 4 shield (0.01) (0.01) (0.01) 0.25 0.15
Biological 0.15 0.1 0.15 0.25 0.4 5 shield (0.01) (0.01) (0.01) 0.4 0.25
13 Biological 0.25 0.2 0.25 0.5 0.7 6 shield (0.03) (0.03) (0.03) 1 0.7
8” gap below 1 0.7 0.7 2 1.0 2.0 7 biological shield (0.5) (0.3) (0.25) 2 (0.3) (0.25)
Upper steel 10 10 10 30 15 20 8 thermal shield (10) (10) (8) 20 (6) (10)
Upper Al 1000 1000 1250 12000 2500 3000 9 thermal shield (800) (800) (1000) 5000 (2000) (2000)
9000 20000 17000 7” gap above 9000 7000 10 top tube sheet (3000) 25000 (8000) of RV (3000) (3000) 17000 (7000)
Top tube sheet 2000 1750 2000 18000 10000 10000 11 of RV (1500) (1250) (1250) 11000 (3500) (3000)
2000 2000 2000 1750 1250 1250 12 RV zone (1500) (1500) (2000) 2000 (1500) (1500)
2250 2500 2500 2000 1250 1250 13 RV zone (1500) (1750) (2000) 2000 (1500) (1500)
2500 3000 3000 2500 1500 1500 14 RV zone (1500) (1750) (2000) 2000 (1500) (1500)
Bottom tube 2250 2500 2500 18000 3000 7500 15 sheet of RV (2000) (2000) (2000) 13000 (3000) (3000)
5” gap below 7500 7000 7500 15000 9000 27000 16 bottom Tube Sheet (3000) (2500) (3000) 31000 (3500) (4000)
Lower Al 1500 1500 1500 7500 2500 7500 17 thermal shield (800) (800) (1000) 10000 (1500) (1500)
14 Lower steel 0.3 0.2 0.25 0.7 0.5 0.7 18 thermal shield (0.5) (0.2) (0.5) 1 (0.5) (0.5)
Main floor 0.15 0.15 0.15 0.25 0.25 0.25 19 plate (0.3) (0.1) (0.2) 0.3
Cross header 0.15 0.15 0.15 0.25 0.2 20 in LHR (0.2) (0.15) (0.2) 0.2 0.2
6.3 Consideration for Functional Systems
Decisions on which systems must remain functional is made during the planning of the transition and are based on: (a) An evaluation to ensure that safety requirements will continue to be met, (b) Support of human entry or occupancy for surveillance and maintenance, (c) Possible use during future phases of decommissioning, (d) Restrictions posed by the current operating licence.
Some considerations may require cost–benefit analysis of: (1) Energy consumption, surveillance and maintenance requirements; (2) Replacement of complex systems with simpler ones; (3) The possible need to achieve a safer state;
7 THE IMPACT OF THE TRANSITION AND DECOMMISSIONING ON HUMAN RESOURCES [4]
There will inevitably be constraints on the approach to staffing during decommissioning. In some facilities staff numbers are likely to be held close to operating levels (e.g. in reactor plants until the fuel has been removed). In other plants the change will depend on the need to stabilize or remove existing hazards. The number of operating personnel needed will eventually fall as shown in (Fig.9). (Fig.9(a)) illustrates the general trend in staffing levels as the facility is shut down, during the post-operation transition, when the facility is placed in a condition of SE, and during final dismantling. The reduction in operating staff as systems are retired and licensing conditions are reduced is somewhat compensated for by additional staff being required for preparation of the SE and final dismantling. (Fig.9(b)) illustrates the trend for a facility proceeding to dismantling soon after shutdown. In this situation, a significant increase in staffing levels is needed for the decommissioning activities. A number of basic points need to be addressed and decisions made concerning the following: (a) The required organization,
15 (b) A staff reduction profile, (c) The use of operating staff to undertake decommissioning project tasks, (d) Sharing of key resources among plants, (e) Policies for choosing which work will be contracted.
Fig.9. (a) Staffing trend during transition, SE and dismantling (b) during the transition to immediate dismantling [1]
Early planning with regard to the timing of final shutdown and the selected decommissioning strategy plays a major role in facilitating the management of personnel relocation and the retention of key staff. The staff reduction profile will depend on the numbers, qualifications and
16 experience of the personnel needed for the actual work to be carried out. Retaining a large number of operating staff will inevitably mean that they undertake decommissioning tasks. They will require retraining in new skills and reorientation of their attitudes towards decommissioning, e.g. system isolation, dismantling, draining and flushing, waste characterization, dismantling and size reduction techniques, etc. It is important to provide appropriate incentives to the remaining staff to work effectively and in a manner that safely maintains the decommissioning programme’s schedule, quality and budget. These incentives may differ from situation to situation and, while seeking to encourage a safe adherence to the decommissioning programme, should encourage staff to strive toward completion of the work and mitigate concerns about future employment [4].
At Cirus, the Operations staff is exclusive to the reactor while maintenance, services, administrative, stores, health physics, security, etc. are common to all reactors / facilities of Reactor Group. The staff reduction was carried out in phases taking into account the changed state of reactor. All personnel were retained till the completion of defueling operations and shipment of spent fuels from the reactor. About 50% operations staff were redeployed to other divisions of BARC after completion of defueling operations and shipment of spent fuels. After draining of process fluids and drying of systems and significant reduction in surveillance requirements, about 50% of the remaining operations staff at Cirus was redeployed to new projects of Reactor Group. The health physics personnel were taken off shift and focused, during normal hours, on pre-decommissioning activities succh as radiological characterization and waste treatment. (Fig.10) shows the operations staff trending during transition from operations to decommissioning.
120
100 99 89 80
60
Manpowrer 44 40 37 26 20
0 2007 2010 2014 2017 2018 Year
Fig.10. Operations staff trending at Cirus during transition from operations to decommissioning
17 8 COST REDUCTION BY RETIREMENT OF SYSTEMS OR RECONFIGURATION
During the transition period, activities are planned and carried out which lead to simplified operation, reduced surveillance and maintenance requirements and lower operating costs. This can be achieved by identifying those plant systems which will become redundant after final shutdown. Further consideration should be given to systems that are needed after shutdown but which are costly to operate and maintain, e.g. the capacity of the ventilation system needed to control contamination in shut down facilities can be greatly reduced. Cost reductions will also take place as a result of changes to technical specifications as the license is amended.
Cost savings can be achieved from reductions in: a) Labor b) Power and fuel consumption c) Consumables d) Surveillance and maintenance e) Regulatory and technical requirements including inspections f) Training g) Recycling of material and components h) Nuclear insurance (not relevant to Cirus reactor)
9 DEVELOPMENT OF TECHNIQUES AND TOOLS
While planning for decommissioning and, in particular, the transition period, it is important to identify whether all the planned tasks can be completed using existing techniques and tools, or if the development of new or the adaptation of existing methods and techniques is needed. It is desirable that this development or adaptation be started during the transition period (in laboratory, mock-up, pilot or full scale) in order to have the decontamination and dismantling techniques available when work commences. The development activities will depend on the chosen decommissioning strategy and selection of the best methods for decontamination, size reduction, dismantling, demolition, waste packaging, etc. A related topic is sampling, characterization and location of contaminated materials and areas where decontamination is required in support of decommissioning. It may also be necessary to test decontamination techniques on selected areas of the plant and its components. Similarly, it may be possible to measure activated samples from a reactor to validate computer estimates with actual measurements in order to optimize size reduction, waste shipping and disposal. The transition period provides the opportunity for any additional sample collection and plant characterization.
In Cirus reactor, after permanent shut down of the reactor, as a part of an IAEA CRP on ‘Establishment of Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactor’, studies were
18 carried out on Aluminium tubes of Cirus reactor vessel for assessment of change in mechanical properties due to irradiation and on Graphite reflector for stored energy due to irradiation. A cutting tool was developed (Fig.11) and a sample from one tube (4” OD, J-07 lattice position) was cut (Fig.12). This tool can be used for sample cutting from tube of reactor vessel at a location about 8m down the operating platform. Extensive trials were carried out before deployment of the tool at site. Radiation field on the sample was 1R/hr. Detailed characterization of the sample is being done. The results of this sample and similar other samples (planned) would give information about radioactivity of reactor vessel. The dismantling and disposal programme would be made accordingly [8].
Fig.11. Schematic arrangement of Fig.12. Sample Cut from J-09 Tube of Cutting Tool RV
Analysis of various Graphite samples indicates that stored energy (Wigner energy) is low in nature and in line with the neutron fluence exposure. There is no concern of uncontrolled release of energy. Graphite plugs for sampling could be removed easily and they were found intact. On inspection inside thermal column, it has been found that graphite blocks are not deformed (Fig.13 & 14). This information will help in planning for dismantling of graphite reflector during pile block decommissioning. It seems that graphite blocks can be dismantled easily and they need not be cut [9].
19
Fig.13. East Thermal Column Fig.14. Hole after removing the plugs
10 WASTE MANAGEMENT [1]
Activities during the transition period have the potential to increase both the volume and the variety of wastes generated. Planning should ensure that there is sufficient capacity for the treatment of these wastes, their storage or transport and disposal. The issues that may need to be considered include: a) The wastes that will arise during the transition period; b) Wastes held in interim storage at the facility which need to be recovered for treatment, conditioning and disposal (e.g. sludges, ion exchange resins, spent radiation sources, scrap components); c) Long term storage requirements; d) Wastes from decontamination and cleanup operations (e.g. additional resins from chemical decontamination, demolition wastes, etc.); e) Availability of disposal routes, including transportation; f) Materials and equipment left over from experimental and research programs; g) Waste retrieval and conditioning methods; h) Waste characterization programs and techniques; i) Waste minimization programs and techniques; j) Clearance levels; k) Regulatory authorizations.
A preliminary estimate of the metallic wastes from various system of Cirus reactor except the reactor core structure, that will be generated during its removal and dismantling has been assessed. The measured radiation dose rate data on various equipments is also available. The solid waste category of the metallic wastes has also been identified and majority of the waste
20 qualifies for Category-I solid wastes and very low volume of metallic wastes qualify as Category-II or Category-III. Data regarding the solid waste that is likely to get generated during the second phase of decommissioning of Cirus reactor involving removal and disposal of peripheral components is enclosed. The (Table.4) gives details such as volume, material composition, expected radiation field, radio nuclides, type of lining etc. Data of in-core components is being worked out. The volume indicated is based on the external dimensions of various pipelines / equipment considered for disposal. Appropriate volume reduction techniques need to be evolved for deciding the final disposal methodology for these waste materials.
Table.4. Estimate of solid waste excluding reactor core structure
System of Component Dimensio Material Expected Volume Radiation Remarks origin ns/ waste (cubic Field & Volume category feet) (mR/h) (Expected Approx. radio- nuclides) Primary Pumps 5 nos. CS I 110 1- 2 mR/h Cooling (3 ft X 3 Water ft) system Shell and tube type 6 nos. CS, I 910 10- 50 mR/hChannel & shell HX (12 ft X 4 Cupro- cover have (Fe-59, Mn- ft dia) nickel, FRP/rubber 54, Mn-56, Silicon lining Co-60, Zn- bronze 65, Eu-152, A1-A-20” pipe line 80 m CS I 600 <0.5 mR/h The underground Ce-144, Sb- A1-B-20” pipe line 160 m CS I 1200 <0.5 mR/h portions of these 125, Cs- A1-C-20” pipe line 160 m CS I 1200 <0.5 mR/h lines are having 137, Ru- external 106) weathering protection coating. A-2A-10” pipe line 360 m CS I 2150 <0.1 mR/h A8A-8”/8B-8” pipe 230 m CS I 275 <0.5 mR/h line A8G-3” pipe line 50 m CS I 10 <0.5 mR/h A8N-6” pipe line 300 m CS I 200 <0.1 mR/h A8M-6” pipe line 40 m CS I 30 <0.1 mR/h A2C-6” pipe line 150 m CS I 100 <0.1 mR/h Small lines (2, 2.5” 200-300 m CS I 25 <0.5 mR/h etc.) Riser pipes 200 nos. SS I 10 5-15 mR/h (2” X 2
21 System of Component Dimensio Material Expected Volume Radiation Remarks origin ns/ waste (cubic Field & Volume category feet) (mR/h) (Expected Approx. radio- nuclides) ft.) Trunion valves 400 nos. SS I 30 <1.0 mR/h (6” X 4”) Smaller pumps 12 nos. CS I 120 2-15 mR/h Big size valves (4” to ~60 nos. CS I 100 <1.0 mR/h 20”) Smaller valves ~500 nos. CS I 10 <1.0 mR/h (¼” to 3”) Upper header 10 ft ring SS I 20 1-3 mR/h dia X 10” dia with 17 cross headers Lower header 10 ft ring SS I 20 2-5 mR/h dia X 10” dia Ion exchanger beds 6 ft X 4 ft CS I 500 10 - 15 The IX beds and (Sy. I – 4 nos., Sy. V dia mR/h associated pipes – 2 nos.) are rubber lined. Delay loop (pipeline) 5 ft dia X CS I 3000 <0.1 mR/h 150 ft Stand Pipe 18 m X 5 CS I 1200 0.5–100 The internal ft dia mR/h surface of Standpipe is having epoxy coating. Heavy Circulating pumps 2 nos. SS I 5 10 - 50 Water and (2’ dia X mR/h Helium 1’ thick) system Supply pumps 2 nos. SS I 5 10- 50 mR/h (2’ dia X (Co-60, 1’ thick) Tritium in Freezer Driers 2 nos. SS Potentially 5 < 2 mR/h
D2O lines) (1½’ X 1’ active dia) Sorber beds 2 nos. CS and SS Potentially 5 < 2 mR/h (3’ X 1’ active
22 System of Component Dimensio Material Expected Volume Radiation Remarks origin ns/ waste (cubic Field & Volume category feet) (mR/h) (Expected Approx. radio- nuclides) dia) Dump header 25 m X 8” SS I 30 < 2 mR/h dia Heat exchangers (3 12’ X 3’ SS with I 275 10- 50 mR/h nos.) dia SS/ Cupro- nickel duplex tubes) Storage Tank # 1 (27’ X SS I 710 100 mR/h 5½’ dia) Storage Tank # 2 (6¾’ X 8’ SS I 360 10-20 mR/h dia) Storage Tank # 3 (6¾’ X 8’ SS I 360 10-20 mR/h dia) Ion exchanger 2 nos. SS I 60 2-10 mR/h (6’ X 2½’ dia) Helium Blower 2 nos. SS Potentially 1 <1.0 mR/h (1½’ X active ½’ dia) Vacuum pumps 2 nos. SS I 15 <1.0 mR/h (2’ X 2’ dia) Pre-cooler 1 No. (3’ SS I 3 <1.0 mR/h X 1’ dia) Cold traps 2 Nos. SS I 7 <0.5 mR/h (4’ X 1’) 4” line 25 m SS I 8 < 1.0 mR/h 2 ½ -3” line 75 m SS I 15 < 1.0 mR/h 2” line 300 m SS I 25 < 1.0 mR/h <2” line 300 m SS I 10 < 1.0 mR/h Valves 300 nos. SS I 10 < 1.0 mR/h Sea water 30” line 100 m MS Potentially 1620 < 0.1 mR/h 30” seawater line system active is rubber/ 24” line 60 m MS -do- 625 < 0.1 mR/h concrete lined. 14” line 10 m MS -do- 35 < 0.1 mR/h Smaller size
23 System of Component Dimensio Material Expected Volume Radiation Remarks origin ns/ waste (cubic Field & Volume category feet) (mR/h) (Expected Approx. radio- nuclides) 8” line 40 m MS -do- 50 < 0.1 mR/h lines are rubber 36” line 10 m MS -do- 250 < 0.1 mR/h lined 4” line 5 m MS -do- 2 < 0.1 mR/h Valves (smaller) 25 nos. CS -do- 50 < 0.1 mR/h Valves (main header) 3 nos. CS -do- 50 < 0.1 mR/h Ventilation Iodine filter housing 1 no. MS Potentially 336 < 0.2 mR/h system (7’ X 6’ X active [Co-60, Eu- 8’) 52, Alkali tank 1 no. MS Potentially 168 < 0.1 mR/h Cs-137 (8’ X 3’ X active (trace)] 7’) Alkali overhead tank 1 no. MS Potentially 144 < 0.1 mR/h (6’ X 6’ X active 4’) Alkali pump 1 no. MS Potentially 1 < 0.1 mR/h (1’ X ½’) active Scrubber 1 no. MS I 250 < 0.2 mR/h (19’ X 4’ dia) TSRS STSRS pumps 2 nos. MS I 16 < 1.0 mR/h (Fe-59, Co- (2½’ dia X 60, Mn-54, 2’ thick) Eu-152) ATSRS pumps 2 nos. MS I 8 < 1.0 mR/h (2’ dia X 1½’ thick) STSRS HX 2 nos. MS I 100 2-10 mR/h (10’ X 2½’ dia) ATSRS HX 2 nos. MS I 65 2-10 mR/h ( 10’ X 2’ dia) STSRS expansion 1 no. MS I 16 < 0.2 mR/h head tank (5’ X 2’ dia) ATSRS expansion 1 no. MS I 16 < 0.2 mR/h head tank (5’ X 2’ dia)
24 System of Component Dimensio Material Expected Volume Radiation Remarks origin ns/ waste (cubic Field & Volume category feet) (mR/h) (Expected Approx. radio- nuclides) Alkali addition tank 1 no. MS Potentially 10 < 1.0 mR/h (3’ X 2’ active dia) Alkali addition pump 1 no. MS -do- 1 < 1.0 mR/h (1’ X 1’) 4” line 30 m MS I 10 < 1.0 mR/h 3” line 20 m MS I 5 < 1.0 mR/h 2½” line 30 m MS I 5 < 1.0 mR/h < 2” line 100 m MS I 2 < 1.0 mR/h Valves 200 nos. MS I 5 < 1.0 mR/h Twin strainer 2 nos. MS I 8 < 1.0 mR/h (2’ x 1½’) PWL Main pumps 2 nos SS I 8.0 <15 mR/h (Mn-54, (2’ X 2’) Mn-56, Zn- Interchanger 2 nos SS I 4.0 5-15 mR/h 65, Sb-125, (2’ X 1’) Ag-110m) Cooler 1 no SS I 2 5-15 mR/h (2’ X 1’) Filter 2 nos SS I 4 5-15 mR/h (2’ X 1’ dia) Ion exchanger 2 nos SS I 5 5-15 mR/h (3’ X 1’ dia) Heater 1 no SS I 36 5-15 mR/h (5’ X 3’ dia) DN tank 1 no SS I 4 5-15 mR/h (2’ X 1½’ dia) Loop cooler 1 no SS I 25 5-15 mR/h (5’ X 2½’ dia) Surge tank 1 no SS I 13 <1 mR/h (4’ X 2’ dia) Make-up pumps 2 nos. SS I 4 5 - 10 mR/h
25 System of Component Dimensio Material Expected Volume Radiation Remarks origin ns/ waste (cubic Field & Volume category feet) (mR/h) (Expected Approx. radio- nuclides) (2½’ X 1’ dia) Catch tank 1 no SS I 36 10-50 mR/h (5’ X 3’ dia) Catch tank auxiliaries (2’ X 2’) SS I 4 1-5 mR/h Decontamination 1 no. SS I 24 10-15 mR/h trolley (6’ X 4’) Main coolant line 2½” X 25 SS I 3 2-5 mR/h m Other lines ¼’ to 1½” SS I 4 - 200m Valves 350 nos. SS I 15 2-5 mR/h Lagging 5 m3 Rockwool Potentially 180 10-50 mR/h active Test section 32’ X 3½” Zircaloy II 3 > 200 mR/h Rods and Outer sheaths 32 ft X Al II 110 2 – 10 R/h Already disposed irradiation 2.5” off. facilities Tray sections 22 ft X Al III 53 2 – 10 R/h (Cr-51, Co- 2.5” 60, Fe-59, SOR sorber sections 12 ft X Boron III 16 20 – 200 Zn-65, Cd- 1.5” Carbide + R/h 109, Mn- SS 54, Sb-125) SOR Barrels 22 ft X 2” SS III 65 20 – 70 R/h Will be cut in to Stringer assemblies 12 ft X 1” SS + Zr III 22 2 – 7 R/h smaller pieces Liner tube 22 ft X SS + Zr III 8 20 – 70 R/h and packed in 1.5” MS drums for PWL tube 32 ft X 4” Al I 15 10 – 50 disposal. mR/h Rupture rod flask 32 ft X 4” MS II 15 200 mR/h Can for SOR cleaning 32 ft X 4” Al I 15 10-20 mR/h Piston ring assembly 22 ft X 3” Al III 8 Will be cut in to Self-serve tray 5 ft X 2” Al III 15 2- 10 R/h smaller pieces sections and packed in Cooling air / water 1.2 m3 Rubber / -do- 45 2 – 10 mR/h MS drums for hoses SS disposal. Instrument Impulse lines from ¼” & ½” Cu / SS I 80 Will be packed
26 System of Component Dimensio Material Expected Volume Radiation Remarks origin ns/ waste (cubic Field & Volume category feet) (mR/h) (Expected Approx. radio- nuclides) tubing different systems tubes in MS drums
Based on the material of construction of the various reactor systems, a list of probable radio- isotopes, their separation method and method of radioactivity measurement along with other relevant information is given in (Table.5).
Table.5. Prominent Isotopes of concern in Cirus research reactor decommissioning
Clearance Measure Sr. Type of Contaminated Elemental Isotope Half life level ment No. radiation system Separation Bq/gm method 1 H-3 12.3 y β, 100 Concrete, graphite, Chemical LSS heavy water system 2 C-14 5600 y β, 1 Graphite, concrete Chemical LSS 3 Cl-36 3.01E05 y β, 1 Graphite, concrete Chemical X spect 4 Ar-39 269 y β, Concrete Chemical LSS 5 Fe-55 2.73 y EC 1000 Coolant systems Chemical LSS 6 Co-60 5.3 y β, γ 0.1 Coolant systems, Nil HPGe Rx, WD System, γ spect Ventillation System 7 Ni-59 76000 y EC 100 Coolant systems, Chemical X spect Rx 8 Ni-63 100 y β 100 Coolant systems, Chemical LSS Rx 9 Mo-93 3500 y X-rays 10 Concrete Chemical X-spect 10 Ba-133 10.5 y β, γ RCB, Concrete Nil HPGe 11 Cs-137 30 y β, γ 0.1 RCB, Coolant Nil HPGe systems, Rx, WD System 12 Eu-152,154 13 y, 10 y β, γ 0.1 Concrete, primary Nil HPGe coolant system 13 Sr-90 28y β 1 RCB, PCW, WD Chemical Beta System counter 14 Y-90 64h β 1000 RCB, PCW system, Chemical Beta (sr-90 sec WD System counter eq) 15 U-238, 235 4.47E09y α 1 RCB, PCW system, Chemical Alpha WD System sepct
27 16 Pu-239- Am- 24,600 y α 0.1 RCB, PCW system, Chemical Alpha 241 432y WD System spect
17 Ru-106, Zr- Short RCB, PCW System, Not Gamma 95, Nb-95, halflife (<1 WD System considered spect, Mo-99, Ag- yr) Beta 110m, Sr-89 counting etc
At Cirus, for characterization of Co-60 and Cs-137 activity in solid waste, a drum scanning system (Fig.15) has been developed and deployed. Solid waste is loaded in 200 liter capacity drums which are assayed for quantification of Co-60 and Cs-137 activity in the waste. Assay is conducted in two stages. Transmission Computed Tomography is used to estimate linear attenuation coefficient as a function of spatial coordinates. Activity measurements are taken around the drum at certain locations using Emission Computed Tomography and an equivalent source distribution is estimated using the attenuation data obtained in the first stage [10].
Fig.15. Solid radioactive waste drum scanning system.
28 11 MANAGEMENT/ADMINISTRATIVE ISSUES This section highlights selected management and administrative issues which were considered during the transition period such as an inventory of hazardous material (including the radiological inventory), the purchasing and spares policy (with a focus on cost reduction), record keeping and interaction with all relevant stakeholders in the decommissioning process. These tasks are treated on a high priority when the transition period is being planned.
11.1 Taking a Comprehensive Inventory of Radioactive and Hazardous Materials [20]
The objective of the characterization of radiologically and chemically hazardous materials is to provide a reliable database on the quantity, types, distribution, and physical and chemical states of these materials. This should include contaminated land. Characterization includes reviewing existing data and calculations, taking in situ measurements, sampling, analysis and undertaking of further calculations as needed. This provides a significant input to the decommissioning planning process and the development of successful implementation plans. Information should be updated on a regular basis to account for waste disposal, material removal, radioactive decay, etc. It is crucial that the database remain available during SE and dismantling. This information will aid decisions for partial or full decontamination, provision of shielding, partial removal of equipment, waste classification, etc. Although a fair amount of data on radio-characterization of materials is available from theoretical studies as well as actual measurements carried out during refurbishment, a detailed analysis and measurement is required for significant input to the decommissioning planning process and the development of successful implementation plans. For that, a dedicated radio-analytical laboratory is being established at Cirus.
11.2 Review of Purchasing Policy and Spares
Expenditures can be greatly reduced if the purchasing and spare parts policy is carefully reassessed. Many purchasing contracts for components, consumables and services are subject to high quality standards related to the requirements of operating plants in the nuclear industry. Also, retirement of systems leads to a reduction in the need for spare parts. Purchasing contracts can be re-evaluated for their applicability and justification in the forthcoming decommissioning phase. Many components and consumables do not necessarily have to meet the same quality standards as required during the operational life of the plant. On the other hand, components already in stock and meeting these standards could be used in other plants and sold as such. It should be emphasized, however, that to demonstrate that these components meet the standards, the full documentation must be in place and the component must be in demonstrably good condition, e.g. kept under suitable storage conditions. The potential also exists for reuse of refurbished components if required quality standards can be achieved. The policy on stocking of components and consumables should be reviewed as well. Component stock size requirements can be reduced in many cases (e.g. where short delivery times are no longer required) and completely lifted in others.
29
In Cirus reactor, procurement of many consumables such as Cuno filters, acid, alkali, regenerative type resin, liquid nitrogen, dry ice, chemistry lab chemicals, Dowtherm-A, etc were stopped. Quantity of procurement of materials such as lubricants, bearings of pumps and fans, etc. has been reduced. Many equipment of drained systems are available as spares for operating systems.
11.3 Records to Support Decommissioning [11]
Prior to shut down and during the transition period, collection of information and records that will be needed to support decommissioning plays an important role. Several challenges related to record related decisions are briefly described here. (a) When to assemble the collection of records and the database: The late creation of a full set of essential decommissioning records may cause difficulties due to reduced availability of time, resources and personnel. As a minimum, it is important that which records will be needed and their location be identified prior to shut down. This should be supported by a suitable records management system. (b) Future retrievability: Record keeping in a deferred dismantling scenario poses long term issues with respect to both degradation and retrievability. In this case, reliance will be completely dependent on records assembled several decades earlier. In particular, any records stored in electronic formats and media need to take into account future changes to systems. Paper or film records are subject to ageing. Record storage systems will have to meet national requirements. (1) How will they be maintained: Questions that need to be addressed include: (i) Which organization will be responsible for keeping records? (ii) Will there be both central and local copies? (iii) Who will have access before they are needed for decommissioning? (iv) What type of database will be used? (v) What are the quality assurance requirements? (2) Which records are to be retained: Selection criteria depends on future needs. A key factor is whether the decommissioning strategy is to be immediate or deferred dismantling. Immediate dismantling is less problematic because the location of much of the information needed will be known and readily available. Selection criteria are -based on: (i) Technical and safety support (radiological and industrial) for decommissioning activities, (ii) Technical and safety support for surveillance and maintenance during SE, (iii) Compliance requirements for statutory and regulatory instruments, including dose and health records, (iv) Historic or social interest, (v) Defense against litigation. (c) Types of record: Application of the above criteria may still require further focus on specific needs, such as inaccessible areas. Certain areas may not have been accessible during normal
30 operations, but workers may need to access these during decommissioning operations. Knowledge of the radiological conditions in these areas, e.g. around the reactor or within the biological shield, will help to minimize occupational exposure during decommissioning. It is also important that the full spectrum of the material characterization and information on the structural condition at shutdown and the end of the transition period are known and recorded. Most of the currently available characterization techniques are suitable for direct electronic recording.
Information was generated during all phases of the design, construction and operation of Cirus reactor, much of which will be useful during the decommissioning phase. Highest degree of priority had been assigned to the documentation and general maintenance of records of all phases of the reactor during its entire history. Such record keeping had been instituted as an integral part of the overall management process for the facility. This practice of good record keeping will be continued during all phases of decommissioning. In particular, records on following aspects shall be maintained during the decommissioning operations:
(a) Site plans, engineering drawings, specifications and process descriptions; (b) Authorisation of the facility including testing, commissioning, operation and modifications, if any; (c) inventory of radioactive waste, including origin, location, physical and chemical characteristics of radioactive waste transferred or disposed from a facility; effluent discharges and environmental monitoring; (d) Safety and environmental assessment methods and associated computer codes; (e) Results of safety and environmental assessments; (f) Data pertaining to QA, Audits and QC; (g) Records of personnel radiation exposure and health history of occupational workers; (h) Incident/accident report and their remedial actions; (i) Training and qualification of personnel related to all processes, stages and phases; (j) Regulatory inspection;
Cirus reactor has a rich history of keeping excellent documentation. During operational stage of the reactor, monthly / quarterly reports on each system which included operational parameters of the system and comprehensive performance review of major equipment of the systems were prepared. An exhaustive monthly and yearly report on the reactor was also prepared. These reports have been preserved for reference. After permanent shut down of the reactor and core defueling, quarterly system reports have been discontinued. However preparation of exhaustive monthly report on the reactor which includes performance of operational equipment / systems is continued. In view of availability of performance reports of various systems, keeping a record of reading sheets of non-operational systems was no more relevant. Such reading sheets after careful review were destroyed. Drawings of all systems / components have been retained and updated in case of any change in the systems / components. (Table.6) gives a list of major
31 documents relevant to decommissioning phase of the reactor and its storage location. At least one hard copy of each document is stored. Soft copy of the documents are stored in a dedicated desk top computer which is not connected to the network. Many documents have been modified as per the existing status of reactor and various systems.
Table.6. Documents and their Storage Location
Sl. No. Documents Storage Location (s) Custodian 1 Operational Cirus Operations Office Decommissioning Supdt., Cirus documents 2 Procedures Cirus Operations Office Decommissioning Supdt., Cirus 3 Drawings Drawing Office Head, Engg, Services Section 4 QA documents Research Reactor Services Head, QA Section, Division, Cirus Operations Decommissioning Supdt., Cirus Office 5 Technical Audit Research Reactor Services Head, QA Section, Division, Cirus Operations Decommissioning Supdt., Cirus Office 6 Regulatory and safety Cirus Operations Office Decommissioning Supdt., Cirus documents 7 Correspondence with Cirus Operations Office Decommissioning Supdt., Cirus other agencies 8 Radiation safety, dose RSO, Cirus Operations RSO, Decommissioning Supdt., records, etc. Office Cirus 9 Event Reports Cirus Operations Office Decommissioning Supdt., Cirus 10 Training, qualification Research Reactor Services Head, SEMTDS documents Division
11.4 Interaction with Stakeholders
Stakeholders typically include: (a) Regulatory authorities: Nuclear safety, transportation, environmental and radiation protection; (b) Local, regional and national governments; (c) The general public: individuals, communities, pressure groups and media; (not applicable for Cirus) (d) Employees; (e) Shareholders; (not applicable for Cirus) (f) Labour unions; (g) Contractors;
32 (h) Waste management organizations; (i) The nuclear industry; (j) National standards groups and professional societies; (k) International organizations (not applicable to Cirus).
The degree of influence of and the priority given to these stakeholders depends on the individual facilities and local circumstances. Stakeholders provide technical, social, economic, environmental, regulatory and legislative input into the process. It is important that these interactions be initiated as early as possible and developed through the transition period. Cirus being a government of India owned facility, many of the issues related to stakeholders is not relevant. Particularly the vexed issue of loss of employment due to shutting down the reactor was not faced. Initially all employees were retained till reactor was defueled and all spent fuels were shifted from site for reprocessing. Subsequently in phases as the reactor systems were curtailed, manpower was deployed to other facilities in BARC. After further truncation of systems and bringing them to a state which required no continuous surveillance, the manpower was further reduced by deploying them to new projects. Adequate manpower was retained at Cirus to operate and maintain the systems required for SE state of the reactor and to take up pre-decommissioning activities.
11.5 Training to Support the Transition [1]
The extent of training of personnel to support the transition depends upon the activities undertaken. If no dismantling or new activities (e.g. POCO) are to be undertaken, training will be specific to the changing conditions of the facility and the differences between normal operations and permanent shutdown. However, training is required for the dismantling of a non- active plant and the introduction of novel techniques for dealing with wastes. Dismantling of non-active plants can be used to train personnel for the future dismantling of active plants. Training material for personnel assigned to a specific facility should be based on the following considerations:
(a) Facilities currently in operation that are going to be held in a shut down condition: much of the knowledge that was required for past outages, maintenance, refuelling, modernization and modification is needed; (b) Facilities which are to be shut down in preparation for SE or dismantling may require the development of skills in such areas as preliminary plant cleanout, waste conditioning and dismantling activities; (c) Facilities that have been out of operation for an extended period and that require inspection to determine whether additional preparatory work is needed prior to decommissioning: training in this situation will require gaining familiarization with the extant conditions.
Some training subject areas for the transition are:
33 (1) Management: Emphasize project as opposed to production/operational management principles. This will involve training in technical, cost and schedule preparations. Training should ensure that management personnel are familiar with the concepts of: (i) Determining the criteria and conditions governing staff reductions, (ii) Amending the safety assessment, (iii) Cost estimation and budgeting, (iv) Complementing the operating organization, e.g. by use of contractors, (v) Change management, e.g. arrangements to deal with staff reductions, etc. (2) Safety analysis: Training should focus on safety issues for a facility that is no longer in operation. It should also address how the safety conditions could be changed and how the requirements for technical specifications, surveillance and maintenance could be reduced. (3) Plant engineering: Training of personnel responsible for operation of systems and equipment should focus on: (i) Shutdown and isolation of systems, (ii) Determination of the required level of surveillance and maintenance as a result of changes in the safety case. The latter should emphasize both cessation of activities or reduction in their frequency, recording any system changes and tagging and identifying systems. (4) Inspection of orphaned facilities: Where facilities have either been abandoned or shut down for an extended period of time, the training of inspectors should highlight structural assessment (building and plant), roof integrity evaluation and identification of radioactive, chemical, electrical, and other physical hazards. (5) Cost estimation: Many of the cost line items for transition are not normally considered during operation. Those responsible for budgeting should become aware of the differences as well as models used for estimating such costs. (6) Waste management: New waste characterization, waste retrieval and conditioning techniques may be developed, for which training is required. (7) Technical and manual work: Training should emphasize implementation of many of the above subjects (for example permanently isolating systems, changes in surveillance and maintenance, new techniques).
12 IMPLEMENTATION OF TRANSITION ACTVITIES
This section deals with actual operations that are normally carried out during the transition period. These include spent fuel removal, draining and drying of circuits and systems, preservation of equipment, waste removal, waste management, removal of components and system management (e.g. reduction or modification of ventilation systems). The removal of combustible materials, radioactive materials and hazardous chemicals will reduce the potential source term for any potential accident and reduce the hazards. It is important to ensure that the above activities are carried out using trained personnel, with appropriate approved procedures and all engineered safety features in place.
34 12.1 Core Management and Removal of Spent Fuel
Removal of spent fuel is a very important step in the decommissioning of reactors. The preferred solution is the early removal of the spent fuel to a storage facility, to a reprocessing plant or to a disposal facility. Benefits of early de-fuelling include decreased radiological hazards, timely implementation of dismantling, downgrading of the operating licence, shutdown of some systems (e.g. cooling water, surveillance), and reduced safeguards requirements. In addition, as long as fuel remains in the fuel storage pools, continuous manning of the unit with shift workers is required, albeit with a reduced number.
Optimum utilization of fuel before shutting down Cirus reactor was one of the major challenges. It was decided to achieve average core irradiation level as high as feasible by December 31, 2010. Had the normal practice of refueling been followed during the approach period to permanent shut down, a large number of fuel rods would have got discharged at low irradiation levels. Therefore a fuel management scheme based upon shuffling of the fuel rods among the core positions was implemented to (i) maximize the irradiation level of fuel rods discharged from the core at the time of permanent shutdown, (ii) minimize the requirement of fresh fuel rods and iii) minimize the number of fuelling operations and thus man-Rem consumption. Preparations for core unloading were started in advance. Storage positions were created in wet storage block (a water tank with appropriate shielding) for temporary storage of fuel assemblies. For air cooled assemblies, similar positions were created in dry storage block. Fuel assemblies were removed with normal mode of coolant recirculation as required by safety considerations. However number of operating pumps (normally four in parallel) for coolant recirculation was decreased gradually from four to one with progressive removal of fuel assemblies. Few assemblies were removed in pumps shut down state when single pump operation was not feasible. During this operation, cooling to fuel assemblies was available under gravity from a DM water storage tank (Ball Tank) in open loop. Plugs were installed in empty positions to keep radiation background low in the working area. Other assemblies such as isotope rods, shut-off rods, experimental fuel assembly in high temperature-high pressure loop; etc were also removed. The whole core unloading was carried out with emphasis i) to minimize waste generation, ii) reduce man-Rem consumption iii) recycle and reuse materials. With proper planning, only about 37% of man-Rem budget was consumed in core unloading [12].
After sufficient decay of gaseous fission product activity and reduction in decay heat, fuel portion of fuel assemblies was cut underwater with appropriate tools and fuel sections were sent to reprocessing plant. Radio-isotopes were delivered for further processing and utilization. Irradiated assemblies were cut into pieces within a specially made Lead enclosure and then sent to waste management division for storage and disposal. Active shielding sections of fuel assemblies were also cut and disposed to waste management division.
35 12.2 System Cleanout Operation It is essential that process and auxiliary fluids from redundant systems be removed and disposed of while personnel are available who are trained and qualified to operate that equipment. After removal, the systems should be flushed until residual contamination is below predetermined criteria and dried as appropriate. The criteria should be based on (a) regulations, (b) an assessment with respect to future decommissioning worker safety, or (c) limiting degradation (e.g. caused by corrosion) while in SE. Organic fluids or hazardous chemicals used during operation, e.g. lubricants, hydraulic oil, acids, etc. are removed and disposed of during the transition period. Radioactively contaminated organic and flammable fluids, as well as non- radioactive hazardous fluids (e.g. PCB transformer oil) or solids (e.g. asbestos) will require special disposal procedures.
12.2.1 Removal of Moderator and Cover Gas
(Fig.16) shows a simplified schematic diagram of moderator and cover gas system. Immediately after shutting down the reactor permanently, moderator (heavy water) from reactor vessel and piping was transferred to storage tank. Subsequently, it was transferred to SS drums and sent for upgradation and utilization in other facilities in BARC. Residual heavy water lying in pipe pockets, instruments and at bends was recovered by drying with Helium (cover gas) and condensing in moisture recovery system. This was done to recover the precious heavy water as far as possible and to bring down tritium activity in the system which will be helpful during dismantling of piping and equipment of the system.
RCU-2 RCU-1
F.D. R.V. SURGE POTS SORBER RELIEF POT BED
V-5030 CV & DV
ST-1 ST-2 ST-3 I.X.
S.P. V-5107
V-5020 H. C.P. EX. GAS HOLDER
Fig.16. Simplified schematic diagram of moderator and cover gas system
Cover gas Helium was replaced with high purity nitrogen after removal of bulk of the heavy water. Pipelines were cut at the lowest elevation in the absence of a flange joint for draining heavy water. In view of certain dead zones such as bottom of tanks, pumps, etc, the system was
36 divided into segments and drying of these segments was carried out by suitable valve manipulations. Heavy water condensate was collected in freezer dryer. Each segment drying was assumed to be complete when dew point of the segment came below -300C. After drying of all the segments in this manner, drying of the whole system was again started. Drying of pipelines and equippment was considered to be complete when the dew point temperature came below -30°C. This was further corroborated by no mmore collection of condensate in the freezer dryer. Subsequently maintaining cover gas was stopped. The space in reeactor vessel, pipelines and equipment is occupied with normal conditioned air. Tritium activity in the exhaust air from reactor which is due to leakage of heavy water moisture laden cover gas, is showing a reducing trend as indicated by the measurement of activity in exhaust air from reactor (Fig.17). Tritium activity in exhaust air came down from about 30000 Bq/M3 during operational phase of reactor to about 600 Bq/M3at the end of completion of drying of heavy water from the system. During April-July 2013, increase in tritium activity is observed as pipelines were opened for draining of heavy water.
Fig.17. Tritium activity (Bq/M3) in exhaust air from reactor.
12.2.2 Draining of Primary Coolant
(Fig.18) shows the simplified schematic diagram of the primary coolant system. After permanent shut down of the reactor, primary coolant system was maintained in preservation mode pending a decision on selection of decommissioning strategy. As part of preservation of PCW lines, they were kept full of DM water. Primary coolant flow from Ball Tank (BT) to Dump Tank #3 (DT #3) was maintained in open loop through 29 nos. dummy rods installed in the pile. Water from DT #3 was pumped back to BT @ 200 igpm using make-up pump. The BT make-up line
37 developed leakage in late 2014 causing gradual reduction in DM water inventory. To reduce rate of loss of DM water inventory, efforts were made to minimize BT make-up requirements. Open loop flow through pile was reduced from 200 igpm to 100 igpm and subsequently to 50 igpm. As the water inventory reduced to a very low level in Ball tank, open loop circulation was stopped which resulted into complete stagnancy of the primary coolant. Activity of the leaked water was BDL. This mode of PCW system preservation was reviewed and the experts opined that keeping the lines drained and dry was a better mode of preservation [13]. Subsequently the lines were drained and DM water was recovered in DT #3 for use. This also resulted into reduced liquid waste generation and processing. Drying of piping was started but subsequently stopped as it was decided to dismantle the pipes and components lying outside reactor structure as part of deferred decommissioning strategy.
Initially close loop recirculation of system was carried out fortnightly to monitor / maintain system chemistry and to monitor leakage from the system. After review of system chemistry for about two years in this mode, fortnightly recirculation was changed to monthly recirculation. System chemistry has been found stable since permanent shut down. Auxiliary circuits of the system were kept isolated. Only one purification circuit was maintained to control water chemistry. Wet Storage Block (WSB); located inside the reactor containment building, was drained and kept dried along with its recirculation system which includes heat exchanger, pumps, pipe lines and mixed bed cartridge. Copper tubing of FFD system have been progressively removed for utilization in Dhruva reactor.
Draining of coolant from system has resulted into significant reduction in surveillance requirements which is a vital component of bringing the reactor to a ‘safe storage’ status. Adequate DM water inventory and quality in Dump Tank #3 is being maintained for meeting make-up requirements of RCB water bays. An alternate provision has been made for the bay make-up from Dhruva reactor.
38 Service water emergency FORWARD Cooling line ECCS VENT
STAND PIPE
UHR I.X. BALL TANK A-2A-10” SEA DEAERATOR WATER D.T. Emergency A-1A-20” cooling
R.V. A-2C-6” H. A-1D-20” A-8A-8” PCW . EX. PUMP FUEL ROD A-1C-20” V-167
LHR DELAY LOOP
FQOV A-1B-20” RAD. ORIFICE Monitors V-1421 RQOV V-356 DT-1 DUMP A-8N-6” TANK-3 DT-2 V-1398 BT MAKE-UP PUMP P-17-24/25
Fig.18. Primary coolant system schematic
Nearly all equipment of this system have become redundant. Equipment lying in the reactor annulus area are progressively being dismantled. DM water requirement has drastically come down and being met with supply from Dhruva reactor. DM water purification plant has been dismantled (Fig.19) and the space has been utilized for installing an ultrasonic decontamination machine (Fig.20). The space was found suitable for the decontamination machine due to its proximity to active liquid waste collection sump. Purification needs of DM water are being met with disposable resin cartridges.
39
Fig.19. DM water purification plant Fig.20. Ultrasonic decontamination machine
12.2.3 Draining of Pressurized Water Loop (PWL) System
High pressure (1500 psig) high temperature (500°F) loop located in one of the 4” size pile positions was used for testing of fuels for power reactors. Subsequent to permanent shutdown PWL system was defueled. It was then drained, washed with DM water and dried with compressed air as part of rationalization of surveillance requirements without affecting the system preservation. The system is primarily made of SS except the zircoloy test section, hence corrosion related degradation is not expected. Preliminary radiation survey on pipelines in Upper Header Room (UHR), Lower Header Room (LHR) and Loop rooms with laggings remaining in place has been carried out. Radiological characterization of activity present in loop coolant was also carried out. Periodic drying of pipelines and equipment was carried out. Dowtherm was removed from Dowtherm system and transferred to Dhruva reactor for use. Some equipment such as safety valves PSV-11 & 12, level& temperature transmitters, high pressure make-up pumps were also removed and sent to Dhruva for utilization. Power supply to all equipment of PWL has been isolated. As no more cooling to various equipment located inside the reactor containment building was required, the loop cooling water pipeline was isolated, drained and terminated before it entered containment building. This was done to reduce leakage hazards. PWL system has been brought to a state requiring no surveillance at all.
12.2.4 Thermal Shields Recirculation System (TSRS)
Aluminium and Steel thermal shields, apart from gamma heat removal, provide shielding against radiation in working areas above and below the reactor. Close loop recirculation of both the aluminium and steel thermal shield system was done fortnightly by operating pumps and system chemistry, leak rate,etc was monitored. The schedule of fortnightly recirculation was reviewed after about two years and was changed to monthly. System chemistry has been found stable. Leakage from Aluminium thermal shield has been found about 3 lpm; almost steady since pre- shutdown period. Subsequently, an estimate in increase in radiation fields in UHR and LHR was
40 carried out in case of draining of the shields and was found insignificant. The shields were drained and dried. This resulted into significant reduction in surveillance requirements and leakage hazards.
12.3 Ventilation System
This system is being operated at normal or reduced flow mode as required. Even at complete ventilation shut down for extended period, air activities in various areas have been found normal. As this system is required till completion of decommissioning of the reactor, all components (except Iodine filters) of the system are being maintained as they were during operational phase of the reactor. However, changes have been implemented to make this system simple wherever applicable to reduce surveillance. During operational phase of reactor, four numbers of trips; namely i) Reactor exhaust air temperature high, ii) High differential pressure across HEPA filters, iii) High Iodine activity in exhaust air and iv) Off-normal reactor hall pressure, used to activate isolation of containment by closure of supply and exhaust dampers. These dampers are ‘air to open’ and ‘spring to close’. The dampers close under spring action for providing quick containment isolation. Provision exists for keeping the dampers on manual mode wherein the dampers are detached from spring and no air supply is required for their operation. Following core unloading and removal of all spent fuels from site, source of Iodine activity and reactor hall pressure off-normal are no more existing. Therefore, these two trips were removed to simplify the control circuit. After removal of air cooled assemblies from reactor as part of core unloading, requirement of operation of air compressors was mainly to keep ventilation system dampers open for operation of fans. This was unnecessarily resulting into power consumption and associated surveillance requirements on the compressors. The dampers were put on manual and normally were kept open. Compressor operation was made need based. The normal ‘manual open’ status of dampers required changes in control circuit of fans for their operation. Direct trip of supply / exhaust fans was introduced in case the other one tripped. Earlier the fans used to trip on closure of dampers.
All components of the system such as dampers, fans, ducts, stack, filter banks, air treatment plant, etc. are being maintained in working condition. Choking of HEPA filters is less at reduced air flow resulting into savings on filter replacement as well as reduction in active waste. Wear and tear of dynamic equipment is also less resulting into less maintenance costs. The 122 m high concrete stack was repaired as it was showing degradation due to ageing.
12.4 Sumps and Liquid Waste Transfer System
There is regular generation of liquid waste due to normal decontamination activities. However there is considerable reduction in generation of liquid waste after draining of coolant from various systems. Considerable liquid waste of very low activity is generated during monsoon due to sub-soil leakage in underground ventilation duct. The liquid waste is collected in various
41 sumps and is subsequently transferred to Effluent Treatment Plant through waste transfer lines. There, after treatment, the waste is discharged to sea. All components of the systems such as sumps, waste transfer lines, pumps, liquid level instrumentation and associated alarms, etc. are being maintained in working condition. The sumps and liquid waste transfer lines are tested for their integrity as per surveillance schedule followed during operational phase of the reactor. Many of the sumps had old pneumatic level instrumentation which was replaced with modern electronic ones in view of the decision to not to operate the air compressors continuously.
12.5 Service Water & Spray Pond / Machinery Cooling Water (MCW) System
Service water is used for drinking and as raw material for DM water production. Spray pond / MCW system is used for condenser cooling of refrigerating machines and other equipment such as compressors, fluid coupling of Reactor Hall supply fan, gland cooling of various pumps, etc. Normal operation of these systems is being continued. Operation of service water system which is common to Cirus and Dhruva reactors has been handed over to Dhruva in view of staff reduction at Cirus. Service water and machinery cooling water lines leading to redundant systems such as DM water production plant, PWL cooling water system, Wet Storage Block heat exchanger, etc. has been terminated at suitable locations to reduce surveillance and hazards due to leakage. (Fig.21) shows such one termination in cable room.
Fig.21. PWL cooling water line terminated in cable room
42 12.6 Mechanical Handling Equipment
Various cranes, chain pulley blocks, monorails are required for handling equipment and are being maintained. Similarly, Vertical flask (Fueling Machine) and fuel transfer buggy to Rod Cutting Building (RCB) is also being maintained for handling active plugs and tubes of core and transfer of suitable active components from reactor hall to RCB under water. All components of the systems are being maintained in working condition and their surveillance is continued as per the schedule followed during reactor operation.
12.7 Rod Cutting Building (RCB)
There are no more spent fuels under storage in RCB. However, it has about 6.5 T of Uranium slurry (U-slurry) of spent fuels which came out from spent fuel rods after failure of cladding due to prolonged storage. The slurry is lying at floor under about 17’ water depth. RCB water bays are made of concrete and are not lined. Bays’ integrity is checked biennially and so far, there is no leakage from the bays. However, in view of ageing of the bays, a plan has been implemented to collect uranium slurry in SS containers (Fig.22) for long term storage under water till it is reprocessed. Bay water chemistry is maintained by its water recirculation and purification system. RCB exhaust air system is also operated to control air activity. All systems of RCB are being maintained. (Fig.23) shows the schematic of U-slurry collection set-up.
Fig.22. SS containers for storing U-slurry Fig.23. Schematic of U-slurry under water collection set-up
12.8 Power Supply System The safety classification of electrical power supply based on its reliability was no more relevant in the permanent shutdown status of Cirus after shipment of all the spent fuel assemblies. As DG sets, Motor-Generator sets, Motor-Alternator sets, Battery Banks, several breakers and supply
43 panels are still available for operation, they are being maintained and operated. High Frequency MA sets used earlier for operating blowers of Helium cover has, were removed and disposed as inactive scrap. Power supply to Pressurized Water Loop (PWL) system was isolated after removal of the test fuel assembly. PWL DG was declared redundant. Subsequently equipment of the systems from which coolant have been drained, were electrically isolated. (Fig.24) shows the trend of electricity consumption after permanent shut down. This helped in reduction of electrical and fire hazards apart from savings on surveillance cost. Aged Motor-Alternator sets were replaced with an available 25 KVA Inverter. This inverter was earlier used to provide power supply to X-ray unit used for inspection of fuel assemblies and had become redundant after removal of fuels. Power supply scheme of trip and alarm panels in control room was simplified. Trip panels were electrically isolated. A few alarms which are needed to monitor important parameters of the reactor and status of important equipment have been moved to a single panel. Equipment which are not required for the current status as well as the status during decommissioning of the reactor, have been isolated. Normal off-site power supply, in general, is sufficient for Cirus except for power supply to fire alarm system and health radiation monitoring and alarm instrumentation which are supplied by Motor-Generators backed up by a floating 125 V DC battery bank. Two 220 KW main reactor DG sets, although about 60 years old, are still available. The DG sets are being maintained to provide power supply to new facilities under installation at site. However, ‘auto’ starting of DG sets is not warranted for the current status of reactor. The DG sets are kept on ‘manual’ and started as per the need. Schemes for further simplification of power supply are being implemented.
1.80E+07 1.60E+07 1.40E+07 1.20E+07 1.00E+07 8.00E+06 6.00E+06 4.00E+06 2.00E+06 0.00E+00 2010 2011 2012 201320142015 2016 2017 Year
Fig.24. Electricity consumption after permanent shut down
44 12.9 Control & Instrumentation Systems
Power control system is no more required and hence it need not be preserved. Instrumentation related to operation of ventilation system and other auxiliary systems, RCB operations and health gamma monitors are being maintained. Some of the process / radiation parameter signals which were in use in the reactor protection system are being utilized for the purpose of status monitoring of corresponding systems / equipment kept in operation. All other reactor trips and alarms, which are not relevant for the safe storage status of Cirus, are not in use now and have been isolated (Fig.25). Alarms on all the parameters relevant to present day surveillance have been arranged in one alarm panel in control room (Fig.26)
Fig.25. Isolated trip and alarm panel Fig.26. Relevant alarms have been put in one panel
12.10 Seawater System
Seawater was used as ultimate heat sink. This system was common to Cirus and Dhruva reactors. Seawater supply to Cirus was isolated by putting a blank on the main supply line. Seawater from pipelines and equipment was drained. During initial period after permanent shut down, this system was continued to be operated by Cirus staff. After staff reduction at Cirus, this system was handed over to Dhruva staff for operation. Dhruva staff was trained for its operation.
12.11 Compressed Air System
Operation of the main air compressors is being continued to meet the truncated requirements of process air and the requirements of service air and breathing air. Operation of only one compressor is sufficient to meet current requirements. Compressed air lines to plant areas where no more compressed air is required, have been isolated to reduce leakage hazards and surveillance requirements. Provision exists to draw compressed air from Dhruva in case Cirus compressors are down.
45 12.12 30 T /day Low Temperature Evaporation Desalination Plant
A 30-T/day desalination plant (Fig.27) which had been coupled with the reactor during refurbishment worked very satisfactorily and was able to meet normal DM water requirements of the reactor. After permanent shut down of reactor, this plant was completely isolated from primary circuit of reactor, drained of coolant and dried with air. Subsequently useful components were salvaged for use in other facilities. The plant was dismantled (Fig.28). Except very little quantity of pipelines connected to primary coolant system of the reactor, almost all components of the plant were inactive.
Fig.27. Desalination Plant Fig.28. Area after dismantling of Desalination plant
12.13 Reduction of Hazards
After draining process fluids from various systems, hazards such as leakage and flooding in local areas, spread of contamination and airborne activity, electrical shocks due to contact with liquid, etc. have been eliminated to a large extent. Fire hazards have been reduced by proper housekeeping and disposal of inflammable materials. Chemicals such as Sulphuric acid (H2SO4) and NaOH which were used earlier for regeneration of Ion exchange columns were disposed by neutralization. Subsequently the tanks containing these chemicals were dismantled (Fig.29 & 30). Dowtherm A which was used as secondary coolant in pressurized water loop system was sent to Dhruva reactor for use. Redundant sections of pipelines of the operating systems such as service water system, machinery cooling water system, compressed air system, etc. were isolated by putting full face blanks. Operational waste was segregated and disposed as per waste category. Many low level active components of fuel assemblies which were earlier recycled were disposed as active waste. Measures for radiation hazard control are being maintained as per normal status and all the installed radiation monitors for this purpose, except the neutron monitors (HNRAs) in reactor hall, have been retained and are being maintained. Rubber stations have been maintained as necessary.
46
Fig.29. Dismantling and disposal of acid Fig.30. Space after disposal of acid and slurry alkali tanks
12.14 Reduction in Surveillance and Stoppage of Round-the Clock Manning of the Reactor After implementation of suitable changes in various reactor systems, hazards were brought down to the minimum. Leakage hazards were eliminated to a great extent by draining process fluid from many systems. Pipe lines which were no more required were cut and isolated at suitable locations. Electrical hazards were reduced by isolating equipment which were no more required. Similarly fire hazards were reduced by disposal of combustible materials from various plant areas as well as isolation of electrical equipment. Liquid waste generation was reduced to the extent possible by diverting inactive waste to storm drains instead of leading them to sumps in the reactor. A comprehensive review of surveillance requirements was carried out and round-the clock manning of the reactor was stopped from 10th July 2017. Prior to that a few alarms and /or monitors were provided in nearby Dhruva reactor which is manned in round-the clock shifts; important ones being fire alarm, water level in RCB bays, radiation level in RCB, stack sump high level alarm and level of Dump tank #2. The last two parameters are relevant during monsoon as large quantity of low active water gets collected in sumps due to leakage in underground exhaust air duct between reactor and stack. Reactor is currently manned during week days from 0800 hrs to 1800 hrs. All jobs are planned in advance and executed only in the presence of Cirus Operations staff. A detailed checklist has been put into force for starting various systems / equipment at the beginning of the day and to stop them at the end of the day. Similarly a detailed checklist is also in force to lock the various areas of the reactor at the end of the day. Cirus security is still manned round-the-clock. Fire alarm and surveillance camera signal are available with security. Electrical staff which is common to Cirus and Dhruva reactor takes care of surveillance requirements of operating equipment
47 13 SAFETY CONSIDERATIONS DURING TRANSITION FROM OPERATION TO DECOMMISSIONING [14] In a nuclear facility undergoing the transition from operation to decommissioning, structures, systems and components are modified and/or retired and their mode of operation may change. Plant personnel should get trained for the new configuration of the systems. It is very important that the operating procedures and drawings be revised accordingly and in a timely manner. Furthermore, as structures, systems and components are changed, modified or retired for decommissioning, quality controls need to be developed and implemented to ensure that: —Plant and system drawings are updated; —System and facility operating procedures are revised accordingly; —Approval and authorization controls are established and documented; —Scheduling and sequencing of systems to be changed, modified or retired are co-ordinated so as to have no impact on the systems and processes required for operations during the decommissioning process.
The following are the major activities carried out during transition from operation to decommissioning and the associated safety considerations:
13.1 Fuel Handling and Storage The handling of spent fuel represents the highest radiological source term and highest heat load of any activity following the permanent cessation of power operations. Accidents involving spent fuel have a high potential to result in high occupational radiation exposure. Additionally, spent fuel accidents could cause the dispersal of radioactive contamination on and off the site, beyond the controlled area of the facility. This will complicate decommissioning and final release of the site for subsequent use and will significantly increase the cost of decommissioning. Core defueling operations and disposal of spent fuels should be completed at the earliest after the permanent cessation of power operations.
13.2 Drainage of Systems The drainage of systems can result in the spread of radioactive contamination to other parts of the facility and systems not intended to be drained. Draining may spread radioactive contamination to other parts of the system such as low points where contamination may settle or can be drained into the drain basin or receptacle. In all cases, the drainage has to be evaluated as to its potential impact on receiving systems, on radiation monitoring procedures that need to be implemented, and on contamination control devices to be installed to monitor for local transitory radiation and contamination levels. Drainage of circuits during the transition period may also generate high volumes of radioactive fluids, which need to be treated. These fluids may require filters to retain more radioactive material than during normal operation. Consequently, the filter dose rate may exceed the handling or transport limits. In some cases, it may be relevant to monitor the filter dose rate and to remove the filter according to a dose rate criterion instead of a
48 pressure loss criterion. Engineering evaluations need to be performed to assess whether: —Partial drainage of systems will adversely impact the functionality or operability of the remaining system; —The drainage process may result in changes in radiation exposures due to a loss of water (i.e. fluid) shielding; —The liquid processing system is of sufficient capacity to handle the large volumes of liquids; —The locations for venting, siphon break and drain path are adequate.
13.3 Cleaning and Decontamination On the basis of operational and decommissioning experience, cleaning and/or decontamination efforts have typically been undertaken to: —Prepare the system or component for final disposal or storage; —Separate mixed waste (radiological from non-radiological, asbestos from non-asbestos and oil from non-oil) to facilitate conditioning, disposal or transport; —Reduce disposal requirements for a particular waste by reclassifying it from a higher to a lower waste category; —Reduce occupational radiation exposure during dismantling activities; —Reduce public exposure during radioactive material transport.
Although chemical cleaning is well understood and utilized periodically, each application needs to be specifically evaluated as to its effects, because chemical reactivity is highly dependent on the specific material exposed to the reactant or reagent. For example, the following facility and system materials have exhibited accelerated corrosion or erosion or degraded integrity characteristics after being in contact with certain cleaning/decontaminating solutions: —Weld filler material and heat affected zones of welds; —Transition pieces between dissimilar metals; —Base metal (exposure due to cladding defects); —Non-metallic or non-ferritic material used in instrumentation and valves.
Cleaning and/or decontamination may spread radioactive contamination to other parts of the system such as low points where contamination may settle. These situations may lead to high radiation levels owing to the formation of radiation hot spots. Radioactive contamination may also be transported to drain basins or filters, purification components and support piping. This spread of radioactive contamination can be a radiological hazard because concentration levels may be higher than those experienced during reactor or facility operation. Therefore, prior to beginning cleaning and/or decontamination activities, the following aspects have to be considered: —Identification and assessment need to be carried out of the possible parts of the system where contamination may settle, as a product of cleaning and/or decontamination activities; —Monitoring and detection devices need to be installed and operated;
49 —Contingency procedures for spills and emergency situations need to be implemented; —Evaluation of the effects of the cleaning and/or decontamination waste products on the normal waste processing system needs to be performed; —Control devices need to be installed to shut down the operation or mitigate leaks;
13.4 Conditioning and Removal of Operational Waste The conditioning and removal or proper storage of operational waste is important during the transition period because it has the potential to adversely affect safe decommissioning. This operational waste includes combustible materials such as rags, wood, oils, plastics, anti- contamination clothing, gloveboxes and other items used during facility operation. It also includes any liquid waste drained from the systems or solid waste generated as part of the transition process. Waste removal operations undertaken during the transition period are normally considered part of the operational activities. These operations may increase the volume and variety of the generated waste. Temporary on-site storage should take into consideration the following aspects: —Response to physical security threats, —Response during radiological or non-radiological facility emergencies, —Fire detection and suppression capabilities, —Facility operator activities and monitoring of system performance, —Safety system operation and availability, —Exposure of workers, —Containment of radioactive contamination by reducing the potential for the spread of contaminants.
13.5 Retirement, Reconfiguration and Planning for the Provision of New Systems During the transition period it may be decided that a number of systems will not be required any longer, some may require modification and others may be needed for later stages of decommissioning. Retirement or reconfiguration of systems will be strongly influenced by the progress of actions to be taken during the transition period. Additionally, new systems may also be required. In general, systems can be categorized as follows: —Those that are required to continue to operate or need to be modified to support decommissioning, —Those to be removed, —Those to be installed to facilitate decommissioning.
13.6 Changes to Confinement Barriers The analysed postulated accidents may change following the permanent cessation of operation of the nuclear facility. After permanent removal of all nuclear and other radioactive material from the facility, containment requirements may also change. There are instances where containment requirements (in the historical sense) no longer apply during the conduct of decommissioning
50 because the containment regulations and design basis requirements were focused on the operational phase. For example, the design bases associated with leak rate testing of a containment may no longer be applicable during decommissioning because there is no design basis accident that could result in a pressure characteristic requiring containment. However, operators of nuclear facilities are not exempt from adequately controlling and preventing the spread of radioactive material (through the use of confinement systems such as gloveboxes, ventilation under reduced pressure, barriers and other engineered features).
13.7 Safety Analysis Formal safety analysis needs to be carried out to justify any changes to confinement barriers or associated systems. Changes to containment barriers could be necessary to simplify the decontamination and dismantling of structures, systems and components, provide consistency with the actual radiological risks present within the facility, or facilitate working conditions or removal of equipment for disposal. Because changes to containment barriers could have a strong impact on safety and on the operability of systems and components, such changes are assessed and justified by safety analysis. The analysis has to take into account both internal risks (e.g. fire, explosion, load handling, and leakage of vessels and systems) and external risks (e.g. earthquake, flooding, aircraft crash, conventional industrial accidents, intrusion and severe climatic conditions), as required by the regulatory body. Special attention needs to be paid to the impact of external risks on the facility.
13.8 Exposure of Personnel During transition phases, areas within containment and/or confinement barriers may be open for access to personnel which were previously secured during facility operation. These areas need to be assessed to ensure that proper atmospheric controls are present to support human activities. Effective radiation monitoring and personnel exposure controls have to be established based on the conditions prevailing during the activity. This takes into account transient radiation levels that could result from the modification or dismantling of structures, systems and components, system flushes and decontamination, or changes to installed or temporary radiation barriers consisting of water, metal, concrete or plastic materials. Furthermore, dismantling or changes to structures and ventilation systems may represent un-analyzed changes in air pathways, which can significantly affect radiation dose modeling. Appropriate controls need to be implemented to account for changes. Installation of additional radiation monitoring devices needs to be considered and personnel need to be trained to recognize that decommissioning has a high potential to increase radiation exposure or to give rise to unanalysed pathways for the release of radiation. To maintain adequate or appropriate levels of leak tightness or control, work procedures and personnel dose estimates need to be established for all activities involving modification of penetrations passing through or entering containment or confinement barriers or their related subsystems.
51 13.9 Considerations for Long Term Storage During the transition period, the likelihood of an unintended or unplanned radiological release is less than during facility operation. In addition, if a majority of the facility systems are in an inactive static condition, the radiological source term continues to decrease with the decay of radioactive isotopes. The safety assessment must take into account these changes, resulting in a decrease in the requirements associated with containment and confinement structures and systems. Specifically, the necessity of containment pressure requirements, elevated release pathways and dilution methodologies may no longer be needed. The removal of such containment and confinement controls offers the opportunity to increase accessibility, improve working conditions, enhance the scheduling of activities and facilitate the dismantling of additional structures, systems and components. This same evaluation could also be broadened in scope to justify changes in methodology and criteria specifications for system operation, maintenance, testing and periodic inspection, leading to further gains in efficiency.
13.10 Other Accidents Possible during the Transition Period Other accidents involving radiation may occur during the transition period that could result in adverse radiological conditions. These accidents could involve solid, liquid or gaseous radioactive waste and the processing, packaging and shipping of such waste. Specifically, rupture of process piping and tanks containing radioactive material may occur. In particular, the likelihood that such accidents occur may increase during the transition period. Also, because the structures and buildings are changing as a result of decommissioning, there is a high probability that new or previously not considered radiological effluent release pathways may be created. These pathways may not be monitored with appropriate instrumentation and alarms to warn of adverse impacts on the environment. Related accidents include, but are not limited to: —Accidents relating to decontamination, such as leakage of the chemical reagent used for decontamination; —Accidents relating to radioactive material handling, such as falling containers and spillage of radioactive material; —Accidents relating to dismantling, such as falling of heavy components; —Loss of high efficiency air filtration; —Leakages of radioactive liquids, and gaseous or solid waste processing system leaks; —Failure of containment or enclosure; —Unauthorized activity.
14 TREATMENT, CONDITIONING, STORAGE AND/OR DISPOSAL OF WASTE DURING THE TRANSITION PERIOD Most wastes generated during the transition period are similar in nature to those produced during plant operation and maintenance. At the end of the operational life of a facility, effort is generally directed at the removal or reduction of any hazard in all areas of the plant to provide a ‘passive’ safe environment during SE. The amount of work to be undertaken depends on the
52 operations that were carried out within the facility and the nature of the hazardous inventory associated with the process, i.e. radiological, toxic or non-hazardous. Such removal or reduction is important for the transition period although historically this has frequently been delayed until the start of dismantling. However, it should be emphasized that if POCO is deferred until the dismantling phase the associated risks remain and are transferred to the future. Methods for assessing the overall requirements for cleanout, both in terms of the need for and extent of such operations, are available [15]. Ultimately, if significant costs or personnel exposure are involved, the decision-making process will be based on the overall net benefit. At final plant shutdown, all waste remaining from past activities is commonly removed from the plant for treatment, conditioning, packaging and storage or disposal. Waste management includes not only process fluids and sludges but also solid waste (e.g. trash, insulation, loose tools) from controlled areas. The latter can comprise a significant number of items, e.g. in research facilities experimental equipment has often remained in the building years after its use. (Fig.31) shows the experimental set-ups in Cirus reactor hall which were dismantled and disposed. (Fig.32) shows shielding sections of fuel assemblies which were cut and disposed. (Fig.33) shows dismantling of lead shielded horizontal flask which was used for handling active assemblies from horizontal storage holes of Storage Block.
Fig.31. Experimental facilities in reactor hall.
53
Fig.32. Shielding sections of spent fuel rods Fig.33. Horizontal flask dismantling
15 DECONTAMINATION OR FIXING OF CONTAMINATION Decontamination after the end of operation helps to reduce occupational exposure during future decommissioning activities. Decontamination may be necessary in the circuits, tanks and containers to remove the activity from inner surfaces, as well as on the surfaces of components and buildings to reduce the potential for airborne contamination. In general, decontamination that is carried out during the transition period is primarily aimed at dose reduction and is not intended for material clearance. Aggressive decontamination methods can often be applied where the systems are no longer needed for operation. The decision whether to decontaminate a nuclear facility (or parts of it) will in general depend on the type of plant, the radionuclide vector/inventory and other constraints such as: (a) The decommissioning strategy selected; (b) The time available; (c) The availability of funds; (d) Individual and collective doses to workers; (e) Liquid and airborne discharges and their radiological impact on the general public and the environment; (f) Industrial safety requirements; (g) Available waste management and disposal options; (h) Workforce availability, including contractors; (i) Reuse of the buildings for other purposes.
Within established constraints, the optimal decision in general is based on a multi-attribute analysis or an extended cost–benefit analysis [16 & 17]. The extensiveness of the decontamination depends on the decommissioning strategy selected. In a delayed dismantling scenario, natural decay reduces radiation and contamination levels in plant systems and components as well as on surfaces and may render some decontamination superfluous. When the need remains after a long SE time, the effect of physicochemical mechanisms during SE may
54 make decontamination less effective, e.g. due to corrosion layers on metals and deeper migration into concrete surfaces. If SE is planned, decontamination will be considered primarily for the areas that will be accessed during the transition period. An alternative in some cases may be to fix contamination in place to reduce airborne re suspension and facilitate access. However, it is important that surface coatings do not overly complicate future decontamination and measurement.
System decontamination may be performed on radioactive systems in order to reduce the general activity level within the systems in preparation for work during the transition period. System decontamination should be carried out while qualified personnel with knowledge of the relevant systems are still available. Various decontamination methods are possible and it is important that the method and decontamination chemicals be chosen with a view towards available waste treatment installations and minimization of secondary waste.
In case of Cirus, no loose contamination has been found on the pipe lines and equipment in accessible areas except RCB. In RCB, old MS railings have been replaced with SS railings for ease of decontamination and also as the old railing were showing signs of degradations. Lower header room and Upper header room; the accessible areas below and above the reactor, were cleaned with water and EDTA / mild nitric acid to remove hot spots. This was done before starting core unloading to reduce man-rem consumption. Pressurized water lines and equipment were flushed with DM water to remove contaminants from the inner surface. Primary coolant pipelines in accessible areas have very low radiation field and their decontamination may not result into appreciable reduction in radiation fields. In view of the chosen strategy of deferred dismantling, the predominant C0-60 activity will further reduce due to natural decay bringing the radiation field further down. Generation of large amount of secondary waste and its treatment is an issue which needs to be handled if decontamination of pipelines is to be pursued. The cost- benefit analysis indicates that internal decontamination of pipelines need not be taken. However, it is planned to make efforts to decontaminate high cost materials (Cu-Ni tubes of heat exchangers) to bring them to exempt level.
The following decontamination studies have been carried out at Cirus:
Heat Exchanger tubes: These tubes constitute a significant quantity (21 t of cupro-nickel material) in terms of waste to be disposed. Radionuclides along with other corrosion products, which are circulating in the primary cooling system, get deposited on the surface of tubes.
Full-scale decontamination is possible only when the corrosion layer (Fe2O3) is dislodged. Two chemical decontamination methods have been tried on Lab. scale. In first method, a one-step reducing formulation containing 2% w/w Na-EDTA, 5% w/w ascorbic acid and 1% w/w hydrazine hydrate was used. A decontamination factor (DF) of 1.5 could be achieved by this
55 method. In second method, a solution of 4% w/w HCl with 0.5% w/w ascorbic acid as inhibitor was used at room temperature to decontaminate the samples. It was seen that within 24-hour period, practically all the radioactivity and corrosion products had come into solution, decontaminating the samples completely.
Fuel channel isolating valves: Experiments were conducted using conventional decontamination reagents like EDTA, oxalic acid, and citric acid to decontaminate the valves. As the valves were removed for decontamination two 2 years after reactor shutdown, most of the short-lived activity had already decayed. The major contribution to the significant radioactivity on these valves came from Co-60 and Cs-137. From the experiments conducted, it is seen that a pre-oxidizing treatment followed by a reducing treatment gives a good decontamination factor of around 20. It is observed that EDTA is a necessary reagent in the formulation to achieve a good decontamination factor. Low concentration of the reagents can be employed, if the temperature of the solution is raised to a higher value. The duration of the treatment can be extended to achieve a higher decontamination factor. The pH value of the solution was appropriately chosen so as to have less corrosion of construction materials, thus minimizing the consumption of ion exchange resins used for trapping radioactivity. The process can be further improved to optimize concentration, temperature and time to achieve better decontamination factors. Gamma spectrometric analysis of samples of the decontaminant solutions after decontamination was carried out. Results are as shown in (Table.7) below. It shows that a DF of 20 could be achieved.
Table.7. Data for decontamination of fuel channel isolating valves [18]
Construction Decontaminant DF Radio Radioactivity material Formulation Achieved nuclides Removed exposed detected GBq(Ci) Stainless steel EDTA +Oxalic acid 5 60Co, 137Cs 2.1 x 102(5.7) Aluminium Pre-oxidation treatment 10 60Co, 137Cs 66.6(1.8) bronze followed by EDTA + Oxalic acid + Citric acid
16 CHARACTERIZATION AND INVENTORY OF RADIOACTIVE AND HAZARDOUS MATERIALS The characterization and establishment of an inventory of radioactive and hazardous materials within the facility involves surveys of existing data, calculations, in situ measurements and/or sampling and analysis. A database can then be established which will provide significant input into the decommissioning planning process and the development of successful implementation plans. With this database, management may assess and decide on various options and their consequences such as:
56 (a) Operating techniques: decontamination processes, dismantling procedures (hands-on, semi- remote or fully remote) and the required equipment; (b) Radiological and industrial protection of the workers, the public and the environment; (c) Waste management, waste classification and disposal options; (d) Discharge authorization; (e) Cost profiles.
At the beginning of the transition period, sufficient information is collected to assess the radiological status of the facility and the nature and extent of any other hazardous materials present. Data collected during this initial characterization period is generally based on information available at the time of final shutdown, including historical operating records. A survey of the extent of contaminated land should be made early in the transition period. As work progresses during the transition period, the objectives of characterization move towards developing more detailed data concerning the physical, chemical and radiological conditions of the facility, including contaminated land. This includes activation calculations, taking and analyzing of samples, as well as in situ measurements of dose rates and contamination to fill the gaps in the available information. Information gathered during these phases serves as the technical basis for work and project decisions, including cost estimates, exposure estimates, risk evaluation, waste management, scheduling and workforce requirements, particularly with respect to radiological exposures. Since characterization requires time, money and dose commitment, it should be optimized to meet the above objectives [6]. As this information should be updated on a regular basis, it is important that the database remain active during the entire decommissioning period.
Radiological characterization studies at Cirus [18]: During Cirus refurbishment, several components of the reactor including piping, storage tanks, sumps and other equipments were accessible and available for extensive characterization, decontamination and radiation surveying. Samples from most of the components were analyzed using high resolution HPGe detectors and the results recorded. This is expected to give reasonable information on the extent of radioactivity present in various components to be decommissioned. Since the data has been collected after 37 years of service, a reasonable extrapolation would be possible with further information during in service radiation surveys. All the data was generated following a reasonable decay period after reactor shutdown which allowed the short lived radionuclides to decay. As can be seen from the data (Table.8) 137Cs is the dominant fission product and 60Co among the activation products with other radio-nuclides like 90Sr, 125Sb, 144Ce, 152Eu, 65Zn, 154Eu, and 11OmAg present in small quantities.
57
Table.8. Typical characterization data of different components [18]
Primary Fuel Primary Primary coolant Hot spots in coolant channel coolant Heat Expansion tank Reactor structure Pipes isolating Exchangers (Stand Pipe) cooling air ducts valves Gross sp. 6.6 5.6x104 1x102 1.5x103 9.1x105 act. (Bq/g) Fission 50 - 90 % 47% 55% 72% -- products Activation 10-50% 53% 45% 28% >99% products Major 51Cr 60Co (42%) 60Co (25%) 60Co (26 %) 60Co (99%) nuclides 137Cs 137Cs 137Cs (22%) 137Cs (32%) contributing 124Sb (22%) 152Eu (15%) 144Ce (22%) to gross 125Sb (16%) activity
Radiological Characterization of Various Systems: During nearly forty five years of CIRUS reactor operation, significant amount of contaminated equipment ranging from a very low level contamination or radiation dose rate to very high contamination or radiation dose rate has been measured and recorded. Very high radiation dose rates of more than 1 Gy/h exists inside the reactor structure at locations above and bottom of reactor vessel. Equipment of Primary Coolant Water System, Heavy Water System, Ventilation System, Pressurized Water Loop System, Waste Disposal System and area and equipment in Rod Cutting Building are some of the area with significant radiation dose rate due to fixed or transferable contamination. The details of the systems/equipments with major radionuclide are given in (Table.9).
58 Table.9. Major Radio-nuclides in Various Systems
System Areas/ Equipments Radionuclides Reactor Upper Header Room Structure Top concrete biological shield 8” gap below biological shield Upper steel thermal shield Co-60, Zn-65, 7” gap above top tube sheet of RV Mn-54, Cd-109 Top tube sheet of RV Cs-137 (in UHR RV zone and LHR) Bottom tube sheet of RV 5” gap below bottom Tube Sheet Lower Al thermal shield Main floor plate Lower Header Room Primary Lower Header Room Coolant Water Valve Room Eu-152, Ce-144, System Upper Pipe Tunnel Sb-125, Cs-137, Orifice Pit Ru-106, Mn-54, Stand Pipe Room (Restricted Occupancy) Co-60 PCW Heat Exchanger Room (Restricted Occupancy) PCW pump room (Restricted Occupancy) Heavy Water Lower Header Room System Heavy Water Heat Exchanger Room (Restricted Occupancy) Traces of Co-60 and 3 Heavy Water Storage tank room H in D2O and (Restricted Occupancy) Helium lines Heavy Water Circulating Pump Room (Restricted Occupancy) Experimental Room PWL Test Section Pressurized PWL line in UHR and LHR Water Loop Loop Pumps Co-60, Ag-110m Loop Cooler Cs-137 (trace) Loop Heater DNRA tank Ventilation 60 “ exhaust duct Co-60 and Eu-152 System Peripheral duct outlet (in traces) Inlet plenum
59 HEPA filter bank room Waste Reactor Hall Sump Co-60, Cs-137, Disposal Outside Active Sump #1 Sb-125, System Outside Active Sump#2 Annulus Sump #2 and Annulus Sump#4
Characterization of soil around the reactor complex: Several of the primary coolant and waste transfer pipes are laid underground but separated from the various utility systems within the plant boundary. To detect leakage from subsoil pipelines and to check migration of radioactivity, a number of bore holes are provided in and around the reactor complex. Water from these bore holes is quarterly sampled to check the radiological status of the environment around the reactor. During refurbishment for inspection and repair of these pipes, soil was excavated. Several soil samples were collected in and around the plant boundary at varying depths. Most did not show any activity; however, at some places soil was seen to have radioactivity at depths from 1 m to 5m below ground. This was attributed to some leaks from the pipelines during the initial days of operation of the reactor. These pipes have since been taken out of service. As a part of their surveillance, all subsoil pipes are pressure tested at periodic intervals to test for leaks. The soil samples collected have clearly identified the areas where activity has been trapped in soil. This information is very useful in monitoring the area through bore hole samples and for eventual clean-up operation during decommissioning. It was seen that 137Cs is the dominant radio-nuclide activity ranging from 56 Bq/g to 1600 Bq/g with traces of 134Cs, 152Eu and 154Eu.
17 PREPARATION OF A FACILITY’S ROOMS AND BUILDINGS DURING TRANSITION During the transition period, access to rooms and buildings in a radiation facility is defined in at least three ways: routine access, no access and completely isolated [1]. (a) Routine access: Most of the rooms and areas of Cirus reactor fall under this category. In these areas, human access for surveillance and/or maintenance can be as frequent as daily or as infrequent as, say, every three months. Industrial safety standards are provided by either temporary, portable or permanent means. Ventilation, lighting and other safety measures are made available, although they need not be operated when the area is unoccupied. Walkthrough routes for periodic surveillance of unoccupied buildings are reviewed for industrial hazards and appropriate protection put in place (e.g. guardrails, warning signs, selected electrical isolators). Contamination and radiation zones are tightly controlled and delineated to prevent the migration of contamination. (b) No access anticipated: These are the areas which contain systems and equipment which may act as sources of radiation fields and contamination. Hence, all entries to these areas are restricted and only with proper authorization. Main outside active sumps, wet storage block,
60 SFSB chimney access well, SFSB water bays are the examples for this category. (c) Isolated: Entry will not be required until demolition begins. Access to pile block components such as graphite reflector, steel, aluminium and cast iron thermal shields, ventilation duct inside pile block and reactor vessel are the examples of this category.
Decisions as to the type of access needed to specific rooms and buildings are closely tied to an evaluation of the surveillance and maintenance requirements. When the surveillance and maintenance routines are determined and the access requirements are decided on, the results will be important inputs to creating the transition end point specifications [1].
18 PROTECTION FROM EXTERNAL OR INTERNAL EVENTS A number of external or internal events may affect a facility. For example, a fire prevention strategy is intended to eliminate fire hazards to the greatest possible extent. Maintenance of good housekeeping standards and emergency access routes are key features in the implementation of such a strategy. Fire detectors are located in potential fire prone areas and the fire alarm is monitored continuously. All plant personnel are trained in basic firefighting skills. Dedicated Fire Services Section of BARC is available all time. Flooding can be a concern as the reactor is located in coastal area. However, reactor and radiation safety are no more jeopardized as the reactor has been defueled, all spent fuels have been reprocessed, and heavy water has been taken out completely from the system and shipped out of the site. Sufficient barriers to stop water entry into RCB water bays are available. An elaborate EOP is available to take care of flooding situation.
19 REMOVAL OF MINOR COMPONENTS Generally, no major dismantling of radioactive parts of a plant takes place during the transition period, depending on the licensing regime. For example, under US regulations, major dismantling activities are defined as any activity that results in permanent removal of major radioactive components, permanently modifies the structure of the containment, or results in dismantling for shipment of components which contain greater than class C waste, i.e. waste unsuitable for routine near surface disposal. Major radioactive components defined by these regulations could include the reactor vessel and internals, steam generators, pressurizers, large bore reactor coolant system piping and other large components that are radioactive to a comparable degree [19]. Examples of decommissioning activities which are considered minor are: (a) Normal maintenance and repair; (b) Removal of certain, relatively small radioactive components such as control rod drive mechanisms, pumps, piping and valves; (c) Removal of components (other than those defined above as major components) similar to
61 those normally removed for maintenance and repair during plant operations; (d) Removal of non-radioactive components and structures not required for safety. This can entail significant amounts of work and include major nonradioactive components such as cooling towers, transformers and control panels. During the transition period, removal of readily movable equipment which is no longer needed can be considered. These items are either: (1) Packaged and disposed of; (2) Packaged after compaction and disposed of; (3) Decontaminated; (4) Directly released without treatment. In Cirus reactor, the inactive / low active / minor contaminated components have been identified for dismantling and disposal in initial year of SE. Pumps, heat exchangers and piping of drained systems are suitable candidates for this approach. Inactive components of Desalination units were dismantled and disposed as scrap. Low active DM water purification system and cover gas purification system (Fig.34 & 35) was dismantled and disposed to Waste Management Division for near surface disposal. Some usable pumps, refrigerating machines, valves and electronic instrumentation have been removed and sent to Dhruva reactor for further use. Acid, alkali tanks were cleaned, dismantled and disposed as scrap.
Fig.34. Cover gas purification system Fig.35. Area after dismantling Cover gas purification system
20 COST OF TRANSITION ACTIVITIES
The costs of transition activities can be significant and a lack of timely funds during the transition period will severely impair the progress of the work. Decommissioning costs, including the costs of transition activities, are categorized in a standardized list [20- 22]. The list, with a focus on the transition period, includes the following groups: (a) Pre-decommissioning actions, e.g. decommissioning planning;
62 (b) Facility shutdown activities, e.g. removal of the spent fuel, system reconfiguration and retirement, decontamination and immobilization of residual contamination; (c) (Limited) procurement of equipment and materials; (d) (Limited) dismantling activities and characterization of radioactive inventory; (e) Waste processing, storage and disposal (including hazardous waste); (f) Site security, surveillance and maintenance; (g) Transition project management; (h) Other costs, including asset recovery.
The prime costs of the transition period activities are related to labour and fuel removal activities, but also include the purchase of equipment and consumables, contract work, etc. The costs are plant specific and dependent on whichever other activities are being pursued on the site. They are also dependent on the schedule chosen for shutdown of the plant and the start of decommissioning. The costs for specific activities within the transition period should be clearly allocated to the operational or decommissioning base costs to establish an unambiguous boundary. Evaluating decommissioning cost according to a standardized list of cost items [20], including the costs of transition activities, would facilitate the comparison of costs for various decommissioning projects and the assessment of cost differences.
A variety of techniques are used to estimate decommissioning costs:
(a) Conventional bottom-up approaches based on the integration of task costs using the project Work Breakdown Structure (WBS); (b) Unit or parametric methods based on feedback from real jobs to adaptively modify unit costs; (c) Computer models.
In addition, if contractors are to be used to perform some or all of the decommissioning, budget estimates and tendered costs can be obtained from them for the relevant activities. In practice, a combination of techniques is likely to be used, but the conventional bottom-up approach based on a detailed WBS is usually the foundation. The more unique the task, the less likely specific experience from elsewhere or generic information from cost databases will be able to shortcut this approach. Factors can be included to account for slower working in radioactive environments compared with estimates from conventional demolition projects. The WBS should be defined as a hierarchy of work to be done, each large work area broken into smaller units of work at the next level down and so on until a sensible stopping point is reached at a specific task level. Costs can then be assigned at that task level in terms of labor, plant and material. Integration of the various tasks up the hierarchy is facilitated by using a unique numbering system or ‘cost code’ to build up estimates in various rolled-up categories as needed for control and monitoring purposes.
63 An initial project cost (the base cost) determined from the WBS can be adjusted by adding allowances for contingencies and risks. Contingency is added to the base costs to reflect an estimating uncertainty dependent on the level of cost definition; for example, low uncertainty typically ±5%, high uncertainty typically ±50%. Contingency assumes that the WBS is adequate in structure (i.e. it reflects the tasks being carried out). The risk budget is an allowance for the planned work (i.e. the WBS) not accurately reflecting the actual work to be done, usually due to unexpected situations such as a delay in a regulatory authorization. An assessment of risk is made by identifying potential deviations, judging their likelihood and estimating the schedule and cost implications. Some risks will have favorable outcomes and some unfavorable, creating a ‘cancelling’ effect that can be investigated using statistical methods such as Monte Carlo simulations to define the most probable outcome in terms of cost and schedule, with ranges related to appropriate confidence limits.
At Cirus as the transition activities were performed by the regular government employees, no specific labor cost could be assigned for transition activities as the government employee salaries were guaranteed otherwise also. However, at the end of the transition activities, limited dismantling activities of a few inactive and some minor contaminated systems was taken up with the help of contractor. On a competitive basis, for about 0.5 million rupees, 30 M3 waste was dismantled. The waste primarily included inactive desalination unit which consisted of rubber lined pipelines and tanks. The minor contaminated waste primarily consisted carbon steel pipelines of 6” size, 2” size SS lines, MS tanks with / without rubber lining. 25 M3 inactive waste was released as scrap for further use while 5 M3 waste was sent to waste management facility for storage. However, the cost estimate for decommissioning is yet to be started.
21 UTILIZATION OF SPACE DURING SAFE ENCLOSURE
In view of deferred decommissioning of reactor structure, dismantling activities in reactor hall will start only after 30-35 years. Reactor hall is supplied with conditioned air and has service utilities such as compressed air, service water, power supply, etc. In view of this, new facilities for processing of I-131 for radiotherapeutic use (Fig.36) and a mock up fuel fabrication facility (Fig.37) for Advanced Heavy Water Reactor (AHWR) project are being installed. In attached laboratory area of the reactor complex, Pneumatic Carrier Irradiation Facility (Fig.38) was decommissioned and the room was converted into a new laboratory for Prognostic Health Management (Fig.39) [23].
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Fig.36. I-131 processing facility at Radio- Fig.37. Mock-up fuel fabrication pharmaceutical Division. A similar facility of AHWR facility under installation at Cirus
Fig.38. PCF, under dismantling Fig.39. PHM Lab. being set-up in PCF room
22 CONCLUSIONS
It is very desirable to take timely action to place a nuclear facility in a safe, stable and known condition as soon as possible after final shutdown. It is important that stabilization and other activities for facilities, systems and materials be planned and initiated prior to the end of operations. Carrying out these activities during the final stages of a facility’s operational phase and during the transition period are beneficial in that the operational capabilities of the facility and the knowledge of personnel are utilized before they are lost. Actions taken at this time pave the way to efficient and cost-effective decommissioning by eliminating, reducing or mitigating hazards, minimizing uncertainty and maintaining steady progress. The fact that Cirus reactor would be shut down permanently by the end of year 2010 was known about two years in advance. Accordingly, preparatory works for maximum utilization of reactor before its shut down and safe and smooth transition from operation to permanent shut down
65 status were started. A scheme for maximum utilization of fuel was implemented. Several proposed experiments related to neutron radiography and fission fragment spectrometry were put on fast tract to complete them before the reactor was shut down. Procedures for complete core unloading were prepared and approved by regulatory authorities. Fabrication of dummy fuel assemblies had been completed before starting core unloading. Empty positions were created in Wet Storage and Dry Storage blocks for storage of spent fuels and other irradiated assemblies.
A deferred dismantling (safe enclosure) decommissioning strategy has been adopted for Cirus. After shutting down the reactor permanently, preparatory activities for the chosen decommissioning strategy have been started. Core has been unloaded and spent fuel and heavy water has been transferred from site. Process fluids have been drained from the systems which are no more required for decommissioning. The systems which are required for decommissioning or for operation of new facilities under installation at site are being maintained in operational mode. Modifications or simplifications of systems as appropriate has been implemented to reduce resource consumption and surveillance requirements. Round-the clock manning of reactor has been stopped. Decommissioning related activities are being carried out in normal working hours on week days. A large number of manpower has been redeployed to other projects. Sufficient experienced manpower to carry out transition activities has been retained at Cirus. Radiation mapping has been carried out to assess radiation field inside core. Detailed radio- characterization of reactor components has been started. A dedicated radio-characterization laboratory is under installation to speed up this analysis. Waste assessment is going on. Documentation for decommissioning has been started. Technical specifications of permanently shut down reactor has been replaced by Technical specifications of deferred decommissioning. Preliminary decommissioning management structure has been put in place. Short and mid-term plans have been made and they are being implemented after necessary approval. Cirus being the first major nuclear facility under decommissioning in India will give valuable experience in decommissioning of similar facilities in future.
The main conclusions of this report are that: (a) Early planning is the key to a smooth transition from operation to decommissioning and will avoid a no action scenario. (b) Planning for transition requires timely allocation of dedicated human, technical and financial resources. (c) Timely implementation of transition activities will reduce expenditures and hazards, simplify waste and material management and help to keep the workforce motivated. (d) Significant cultural and organizational changes will occur during the transition from operation to decommissioning and need appropriate consideration and management. (e) The availability of relevant data and records is essential for smooth progress into and implementation of decommissioning. A database containing all relevant data needs to be
66 established and maintained. This database should be kept up to date throughout the lifetime of the facility. (f) Implementation of transition will require comparable management focus and workforce attention to detail as during normal operation. (g) Good communication and involvement of all relevant stakeholders is essential for a successful transition from operation to decommissioning.
ACKNOWLEDGEMENTS
The mammoth task of transition from operation to decommissioning of Cirus reactor was a group activity in which apart from various divisions of Reactor Group, several other divisions of BARC also contributed. Particularly the authors acknowledge contribution of Shri R.C. Sharma, Ex. Group Director, Reactor Group for guidance during initial planning and core management and of Shri N. Ramesh, Ex. Reactor Superintendent, Cirus and Shri Alok Srivastava, Ex. Asst. Reactor Superintendent, Cirus for implementation of many pre-shut down activities. The authors acknowledge contribution of Dr. P.V. Varde, Head, Research Reactor Services Division and Shri H.G. Gujar, Head, Engineering Services Section for providing engineering support. The authors acknowledge contribution of Shri P. Sumanth, Head, Research Reactor Maintenance Division, Shri Alex Mathew, MS (Electrical), Shri O.P. Ullas, MS (Mechanical) and their colleagues for implementing suitable changes in various systems. The prime activity of providing health physics coverage and radioactivity analysis was carried out by Shri Ranjit Sharma, RSO and Shri K.S. Babu, Heatlh Physicist. Needless to say, this mammoth job could not have been finished without active contribution of Cirus Operations staff and other technical staff of Reactor Group. The authors would fail in their duty if they don’t acknowledge valuable contribution of each member of Cirus Decommissioning Task Force; particularly Shri Diwakar Mathur and Shri R.K.B. Yadav and Apex Committee of Cirus Decommissioning. The authors acknowledge contribution of Shri Kunal Chakrovarty, Reactor Superintendent, Dhruva and Dhruva Operations Staff for sharing resources and monitoring a few vital parameters of Cirus during off hours when Cirus Operations staff is not at site.
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