PLANNING AND MANAGEMENT OF REFURBISHMENT WORK OF CIRUS

Transcription

1 PLANNING AND MANAGEMENT OF REFURBISHMENT WORK OF CIRUS S. SANKAR Reactor Group, Bhabha Atomic Research Centre, Mumbai, India Abstract The 40 MWt research reactor Cirus located at Bhabha Atomic Research Centre, Mumbai is in operation since the year The reactor operated with an availability factor of around 70% till the year 1990 whereafter its availability factor started coming down due to frequent breakdowns on account of ageing. Systematic ageing studies were therefore carried out to assess the condition of its structures, systems and components. Based on these studies, the refurbishing requirements were identified. A detailed plan was developed, after discussions in the Station Operations Review Committee, for implementation. After taking necessary approvals from safety authorities, the reactor was shut down during October, 1997 for implementing the planned refurbishment activities with an overall schedule of two and half years. It is expected that with the refurbishment, the useful life of the reactor would be extended by about ten years in a cost-effective manner. The paper describes in detail, the various refurbishing activities that are in progress. 1. INTRODUCTION Cirus, a 40MWt tank type, natural uranium fueled, heavy water moderated and light water cooled research reactor with a maximum thermal neutron flux of 6.5 E13 neutron/cm 2 /sec is located at the Bhabha Atomic Research Centre, Mumbai, India. The reactor was commissioned in 1960 and has operated successfully for over 37 years. The reactor core is housed in a cylindrical aluminium vessel with aluminium lattice tubes located between the top and bottom tube sheets of the vessel. The fuel assemblies are loaded through these lattice tubes and moderator heavy water is filled inside the reactor vessel. The reactor vessel is surrounded by two annular rings of graphite reflector, cast iron thermal shields and a 2.5 metre thick heavy concrete biological shield. On top and bottom of reactor vessel, there are aluminium and steel thermal shields with removable concrete biological shields placed at the top (Figure 1). The fuel is cooled by demineralised light water circulating in a closed loop with the coolant flowing through fuel assemblies from top to bottom. Heat from the primary coolant is rejected to seawater in shell and tube type heat exchangers with seawater flowing on the tube side in a oncethrough mode. Shut down cooling is provided by one-pass gravity assisted flow of water from a concrete storage tank (commonly known as Ball tank) of 3.8 megalitres capacity located at a higher elevation than that of the reactor and connected to the system through a set of check valves. Coolant outlet from the core is led to an underground tank through a set of quick opening valves located in the primary coolant outlet line (Figure 2). Cirus operated with an average availability factor of around 70% till the year Thereafter its availability started coming down due to frequent break downs on account of ageing, which necessitated extra efforts for maintenance (Figure 3). Detailed ageing studies were therefore carried out to assess the condition of its structures, systems and components. 1

2 Fig. 1 Based on these studies and performance review of the reactor and auxiliary systems, the refurbishing requirements were identified and a detailed refurbishment program was prepared for implementation to extend the useful life of the reactor in an economic manner [1]. 2. REFURBISHMENT PLAN Initially a preliminary refurbishment plan was prepared and was discussed in detail in the plant operations review committee consisting of representatives from various units such as operation, maintenance, quality assurance and audit, technical services, training, radiation hazards control, reactor physics and chemistry to draw up the detailed plan. All O&M procedures and instructions were discussed in detail in this committee before formal approval was given for the refurbishment works for implementation. Services of experts, where required, was taken to have the necessary technical inputs to optimise the work procedure. Required approvals were also taken from regulatory authority. For timely implementation, the plan is being periodically reviewed by a task force consisting of representatives from various implementing agencies. All works are carried out through a well-established work permit system appropriate controls for radiological protection and industrial safety. 2

3 Fig. 2 The reactor was shut down during October 1997 for implementing the planned refurbishment activities with an overall schedule of two and half years. Following are the major activities involved in the refurbishment work. Core unloading Radiation field measurements on in-core components and plant areas In-service inspection of reactor vessel tubes Repairs to known leaks in primary coolant system piping Pressure testing of subsoil piping of primary coolant system Replacement of primary coolant heat exchangers Repairs to leak from the central inspection shaft of the concrete ball tank Repairs to helium flange joints in the reactor structure region Repairs to inlet coolant line of top aluminum thermal shield 3

4 80 CIRUS - Availability Chart 80 Availability Factor (%) Year 0 Fig. 3 CIRUS availability chart Repairs/replacement of identified leaky sections of helium cover gas piping Thermal safety assessment of graphite reflector Refurbishing of reactor hall conditioner and ventilation system Seismic assessment of safety related concrete structures. Refurbishing of the radiation shielding window of the hot cell facility Replacement of process system piping such as machinery cooling water, chilled water, service water, compressed air, etc. Fire safety improvements Data collection for preparing preliminary decommissioning plan for future decommissioning Utilisation of reactor waste heat for desalination. 3. CORE UNLOADING After shutting down the reactor, the core was completely unloaded of all fuel assemblies, experimental assemblies, isotope tray rods and primary shut down devices. All empty reactor positions were plugged with long aluminium shielding plugs both at top and bottom to bring down the radiation fields in the working areas. About 30 numbers of dummy fuel assemblies were installed to facilitate circulation of water through the primary coolant loop and for control of the water chemistry for system preservation. Moderator heavy water was drained from the system and kept locked up in storage tank. All other process and ventilation systems were kept in normal operation. 4. RADIATION FIELD MAPPING After core unloading, extensive radiation field measurement on reactor in-core components were carried out during November January 1998 i.e. after about 7 to 15 weeks of reactor shut down. Field measurements were taken in 46 pile positions at 20 different elevations between Upper Header Room (UHR) and Lower Header Room (LHR). Radiation fields measured on reactor structural components in three typical positions are indicated in Table 1 [2]. 5. IN-SERVICE INSPECTION OF REACTOR VESSEL TUBES In the year 1971, one of the reactor vessel tubes developed leak and was plugged. Further checking and analysis carried out indicated that failure was random in nature and not of generic type. 4

5 Visual inspection with the help of boroscope was carried out on a number of lattice tubes. An eddy current absolute coil probe was developed to check the overall wall thickness of the lattice tubes. All the tubes were checked and the observation revealed that there was no significant thinning of the tubes. During 1994, one more lattice tube developed leak and this was also plugged. A test probe employing differential-coil (dual frequency) eddy current technique for spotting and assessing local defects in the lattice tube was developed in-house and was put to use after laboratory trials with a defect reference standard. So far one third of the core has been inspected and results indicate that the tubes are generally in a healthy condition [3]. 5

6 6. PRIMARY COOLANT PIPING Major portion of primary coolant system piping of around 1100 metres length is of seamless carbon steel conforming to ASTM- A53 with diameters of pipes ranging from 50 mm to 500 mm. About 70% of piping is laid subsoil and individual sections of piping are joined with mechanical couplings having elastomer seals called "Dresser Coupling". In the 500 mm dia. lines, over 20 such couplings are provided. One such coupling on the reactor coolant outlet line located in a sub-soil concrete chamber (5 m 3.5 m 4 m high), known as Orifice pit, developed a leak some time back. Replacement of the leaky elastomer seals of the joint was carried out without draining the water from the system to prevent the elastomer seals of the remaining dresser couplings developing leak under dry conditions. For this purpose, the concrete chamber was flooded with water and the seals on the coupling were replaced underwater with the help of divers. One observation during the refurbishing outage was development of a crack adjacent to a weld in one of the coolant outlet cross headers made of SS 347 material. In-situ metallographic examination was carried out by transferring the micro-structure to a back reflecting plastic replicating strip. It was revealed that the area surrounding the crack had sensitized micro-structure. The crack propagation was seen to be inter-granular. There are 13 such welds in this cross header and all these welds were scanned ultrasonically after dismantling the header from its location. It was observed that all other welds are free of defects except the one, which had failed [4]. The cross header was taken up for repair, and a section of the header was cut near the cracked region and was subjected to further detailed inspection. This revealed that near the crack a repair work on this header has been carried out during initial installation. Preparations are under way to carry out the required repair using a qualified procedure. 7. PRIMARY COOLING WATER STORAGE TANKS The spherical concrete water storage tank (Ball tank) of shut down cooling system had developed a small leak some years back at a concrete pour joint in the central inspection shaft (Figure 4). During the current outage, this concrete tank was taken up for repair after completely draining the water. Special temporary ventilation arrangements were engineered for safe working in the confined space of the tank. The repair work was carried out from the wet side. The Ball tank and the underground concrete storage tanks which receives shut down cooling water outlet from the core were evaluated for seismic qualification. The design stresses were checked for the design basis earthquake ground motion considering dead weight and live load and a dynamic and static stress analysis was performed to assess the safety margins. The structures have been found safe for present conditions of earthquake loading [5]. The health of the concrete of the Ball tank was also assessed by ultrasonic pulse velocity, corrosion activity, carbonation depth and rebound hammer tests and by testing the core samples. The test results indicate the structure to be in a healthy state. 8. REPAIRS TO CORE COMPONENTS One of the challenging jobs during the present outage is the repair of leaky flange joints of the helium cover gas system located in the reactor structure region. There are eight nos. of flange joints between the aluminum extension pipes connected to the reactor vessel and the stainless steel system piping. These joints are located in the 200 mm vertical gap between the upper steel thermal shields and concrete biological shields. The flange joints are of tongue and groove design with elastomer seals. In order to avoid removal of the large structural components above the flange joints, a remote repair procedure was developed. The procedure involves tightening the flange joints using clamps and has been tested successfully on a full-scale mock-up. Closed circuit TV is used for remote viewing 6

7 Fig. 4 and the tools and clamps are maneuvered into place using multiple ropes. The ropes are manipulated from the operating platform located about 5 m above, somewhat like in a puppet show. Analysis has been carried out, to identify the correct sequence and the extent to which these flange joints could be tightened safely. The repair method was earlier qualified, in a mock-up, through correcting leaks by tightening a similar flange joint in a mock-up with very old embrittled gaskets in place. 9. REPAIR TO TOP ALUMINUM THERMAL SHIELD In Cirus, immediately above and below the reactor vessel, aluminum thermal shields are installed. These shields are cooled by demineralised light water. Each thermal shield is made of two tube sheets inter-spaced with lattice pipes and is of welded construction. The top Aluminium thermal shield is provided with two nos. coolant inlet lines and two nos. coolant outlet lines at diametrically opposite locations. During the year 1995, water leak was observed and was identified to be from the aluminum thermal shields in the reactor structure region. Visual examination with the help of a boroscope through lattice positions indicated that the leak to be from top aluminum thermal shield. This was confirmed by pressure testing. Further, investigations indicated that the leak is from the welded joint between one of the inlet coolant pipe and top tube sheet of the thermal shield. The leak rate was estimated to be about 20 ml/min. The leaky zone being around 5 m below the operating platform and located in-between the steel thermal shield and aluminum thermal shield, any direct repair work would need dismantlement of reactor structure from the top and hence detailed preplanning. As the top aluminum thermal shield has two cooling water inlet lines, a possibility of plugging the leaky line at the leaky zone was considered. Towards this, an analysis was carried out to check coolant flow with only one coolant inlet line in use and was found to be adequate for the design power 7

8 level. A plan has been worked out to plug the line in the leak zone remotely after conducting full-scale mock-up trials. 10. GRAPHITE REFLECTOR In order to assess the thermal safety of the irradiated graphite reflector, samples were taken from the reflector to measure the Weigner stored energy, degradation in thermal conductivity and Wigner energy release behavior over a temperature range from room temperature to 720 o C. The Wigner energy release characteristics of irradiated graphite was evaluated by differential scanning calorimetry (DSC). In addition, planned experiments were carried out after safety evaluation at different power levels prior to reactor shutdown to measure the graphite temperature transients following a reactor trip from power operation and simultaneous reduction in ventilation cooling of the reflector. Based on the results of these experiments and the laboratory measurements carried out on the graphite samples, the stored energy in the reflector is found to be of no significant safety concern [6]. 11. UTILIZATION OF REACTOR WASTE HEAT FOR DESALINATION BARC has recently developed a low temperature vacuum evaporation (LTVE) process for desalination of sea water and a unit of 30 Te per day capacity has been built in-house. Since high power research reactors produce significant quantities of low temperature heat (energy), a scheme was evaluated to integrate the desalination unit with Cirus primary coolant systems such that the technology of utilising reactor waste heat for desalination through LTVE can be practically demonstrated [7]. During the refurbishing outage, the LTVE unit for desalination of sea water will be coupled with the primary coolant system of the reactor. For this, about 7 % of the gross primary coolant flow would be diverted through a plate type heat exchanger working in parallel with the existing primary coolant shell and tube heat exchangers. The reactor waste heat would be utilized for desalination through an intermediate coolant circuit. The pressure of intermediate coolant in the plate type heat exchanger will be maintained at a value higher than that of primary coolant water to avoid ingress of radioactive primary coolant water into the intermediate circuit in the event of a leak in the plate type heat exchanger. This will preclude any possible release of activity to the environment. The system changes proposed would be carried out in a manner that normal operation and operating parameters of the primary coolant systems, will not be affected in any adverse manner. It has also been checked that the safety of the reactor cooling is not affected for postulated design basis events. The final product water from this system would be used to meet the demineralised water make-up requirement of the reactor. The product water quality is expected to be with seawater salinity of 20 ppm. 12. CONCLUSION Some of the important refurbishment works of the 40 MWt Cirus reactor are detailed in this paper. It is expected that after refurbishment, the reactor would be put back into operation with improved safety and reliability and the useful life of the reactor would be extended by over ten years in a cost effective manner. The experience of refurbishment will also aid in planning and execution of decommissioning of the reactor at a future date. 8

9 ACKNOWLEDGEMENTS The author wishes to thank Shri S.K. Sharma, Director, Reactor Group, BARC, for many helpful guidance and discussion in the preparation of the paper. The author is grateful to Shri M.R. Ranade, Head, Reactor Services and Maintenance Division and his colleagues for providing detailed information on the refurbishment work in progress. The author also wishes to thank Dr. P.V. Varde, Shri K. J. Vakharwala and Smt. Veena Ramanathan for helping him in preparation of the paper. REFERENCES [1] V.N. ACHARYA, Ageing Studies and Identification of Refurbishing Requirements for the Primary Coolant System of Cirus, IAEA Topical Seminar on Management of Ageing of Research Reactor, Germany, May [2] S.SANKAR et. al., Generation of Database for Future Decommissioning of Cirus IAEA CRP on Decommissioning Techniques for Research Reactors, RCM held at Mumbai, February 1998 [3] M.R. RANADE, et. al., ISI of Reactor Vessel Tubes of Cirus and Process Water-Sea water Heat Exchangers of Dhruva, IAEA CRP on Application of NDT and ISI to Research Reactors, RCM held at Prague, November [4] K. C. SAHU, Failure Investigation on SS 347 cross-header of Cirus, an internal report of BARC, Mumbai, December [5] M.K.AGARAWAL, et. al. Seismic assessment of the overhead storage tank and dump tank of Cirus, an internal report of BARC, Mumbai, March [6] B.K. DATTA, et.al. "Modelling of CIRUS graphite reflector with stored energy: Comparison of blind predictions with high power measurements", 15 th International Conference on Structural Mechanics in Reactor Technology, Seoul, Korea, Aug [7] K.SASIDHANRAN, et. al., Utilisation of Cirus waste heat for desalination of sea water, an internal report of BARC, Mumbai, January,