- Email: [email protected]
Research reactor KAMINI
Nuclear Engineering and Design 236 (2006) 872–880
Research reactor KAMINI S. Usha ∗ , R.R. Ramanarayanan ∗ , P. Mohanakrishnan, R.P. Kapoor Indira Gandhi Centre for Atomic Research, Kalpakkam, India Received 11 March 2004; received in revised form 26 September 2005; accepted 29 September 2005
Abstract KAMINI is a 30 kW, 233 U fuelled research reactor located at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. The reactor functions as a neutron source with a flux of 1012 n/cm2 /s at the core center. KAMINI uses plate type of fuel in a reactor tank. Demineralized water is used as moderator, biological shield and coolant. Reflectors are made of beryllium oxide with Zircaloy-2 sheath. Cadmium is used as the absorber material in the safety control plates (SCP) provided for power control and shutdown of the reactor. There are three beam tubes and three irradiation locations within the reactor tank facilitating neutron radiography of radioactive and nonradioactive objects, sample irradiation for neutron activation analysis, physics and shielding experiments. The reactor attained first criticality in October 1996 and nominal power of 30 kW in September 1997. This article covers the design description, facilities available for experiments and their utilization for research and development. Seven years of operating experience of KAMINI, which is the only reactor using 233 U as fuel, are also highlighted. © 2006 Elsevier B.V. All rights reserved.
1. Introduction
2. Basic design features
KAMINI (KAlpakkam MINI) is a 30 kW, 233 U fuelled, demineralized light water moderated and cooled, special purpose research reactor located at Indira Gandhi Centre for Atomic Research (IGCAR). Beryllium oxide (BeO) is used as reflector and cadmium is used as absorber material in the safety control plates (233 U fuelled reactors have been operated earlier in many countries, notably, the Shippingport light water breeder reactor in US). The reactor was designed and built jointly by Bhabha Atomic Research Centre (BARC) and IGCAR. Salient features of KAMINI are given in Table 1. The reactor functions as a neutron source with a flux of 1012 n/cm2 /s at the core center and facilitates carrying out neutron radiography of both radioactive and non-radioactive objects and neutron activation analysis. Facilities are also available in the reactor for carrying out radiation physics research, irradiation of large samples and calibration and testing of neutron detectors (Ramanarayanan et al., 1999).
A judicious choice of design features, layout and control philosophy has been made using the experimental results and experience of similar reactors to provide operational flexibility and to achieve inherent safety. That is, in case of any power excursion due to inadvertent addition of reactivity, the reactor shuts down due to negative void and temperature coefficient. KAMINI is a tank type reactor using enriched uranium and aluminum plate type fuel in a light water pool. Physics of 233 U fuelled system including a full scale mock up of KAMINI core has been studied extensively in PURNIMA reactor at BARC to establish the neutronic parameters and safety (Pasupathy et al., 1990; Radkowsky, 1986; Dietrich, 1958). KAMINI is a reflector moderated reactor where 50% of fissions are due to reflector returned neutrons. Zircoloy-2 canned BeO is used as reflector owing to its high reflection efficiency resulting in lower fuel inventory. Choice of Zircaloy-2 sheath for BeO is due to its superiority over aluminum with respect to neutron economy and structural considerations. Fully reflected reactor concept (reflectors are arranged on all sides around the core) is adopted to minimize fuel inventory and to prevent inadvertent addition of large positive reactivity by reflector movement.
∗
Corresponding authors. E-mail addresses: [email protected] (S. Usha), [email protected] (R.R. Ramanarayanan). 0029-5493/$ – see front matter © 2006 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2005.09.033
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880 Table 1 Salient characteristics of KAMINI Reactor type Nominal power (kW) Core size (mm) Fuel Fuel inventory (g) Subassemblies Plates per subassembly Gap between fuel plates (mm) Plate size (mm) Subassembly size (mm) Reflector Number of modules Number of adjustable reflectors Absorber material Excess reactivity ($) Effective delayed neutron fraction ($) Reactivity compensation Moderator/coolant/shield Quantity of moderator (t)
Open tank type (2 m diameter, 4.19 m height) 30 204 × 204 × 275 233 U
(20%)–Al alloy plates with Al clad 600 9 8 6 2 × 62 × 260 66 × 66 × 275 BeO (200 mm thick) encased in Zircaloy-2 sheath 20 3 Cadmium sandwiched in aluminium 0.76 1 (βeff = 0.0033) By reflector addition Demineralized water (pH: 5.5–7.5, K ≤3 S/cm) 12.5
873
Possibility of reactivity addition by the sample, through the pneumatic fast transfer system used for study of short lived activity is eliminated by the restriction on sample volume and strict administrative controls. The reactor core is cooled by natural convection. Any sudden power increase (as in an accident) results in a rapid increase of temperature in the vicinity of the core. This causes shutdown of the reactor due to high negative void and temperature coefficient of reactivity. The reactor vault is located in the basement of the hot cell, located at the ground level to enable lowering of irradiated fuel in front of the beam tube directly from the operating floor of the hot cell. Floor of the reactor vault is 1.5 m below the floor level of the basement area and the core top is 300 mm below the basement floor level with 3 m of water column above the core in the reactor tank, ensuring submergence of the core in case of gross leak in the reactor tank. The building housing the reactor is located at a flood safe elevation and double isolation valves are provided on the demineralized water supply line from outside to prevent accidental flooding in the basement. Stainless steel is used as the material for the reactor tank and coolant system piping with welded joints to minimize the possibility of leak. 3. Design description
The inherent safety of KAMINI reactor arises from the negative temperature coefficient (−0.017 $/◦ C) and the negative void coefficient (−0.0230 $/ml), which act as shutdown mechanisms in case of insertion of excess reactivity. A small inventory of 233 U (approximately 0.6 kg) used in a compact core of about 10 l limits the radiological hazard potential during accidental conditions. As indicated in the safety analysis, for uncontrolled withdrawal of the safety control plate, the maximum power reached is only 115 kW, which can be easily removed by the reactor tank water with the water temperature not exceeding 70 ◦ C without any safety consequences (the reactor safety was evaluated using SECMOD code) (Singh et al., 1980). Physics design with the H/233 U ratio of around 320, being in the under moderated region, ensures that moderator temperature coefficient is negative providing operational stability. Reactivity insertion rate limited to 0.009 $/mm allows enough time for preventive action in case of loss of regulation. Total worth of absorber plates is such that adequate shutdown margin (more than 3.030 $ with one safety control plate withdrawn) is provided while allowing a reasonably rapid startup for experimental purposes. The excess reactivity is provided to fully compensate for the essential operating loads namely, temperature and sample in the irradiation locations but only partial compensation is provided for xenon poisoning. This limits the excess reactivity and acts as self shutdown mechanism if the reactor is operated for more than 10 h at 30 kW power. Excess reactivity of less than 1 $ (βeff = 0.0033) ensures that the reactor cannot attain prompt criticality. Reactivity compensation for burn up loss by reflector addition minimizes handling of fuel. There are no vacant locations in the core, which prevents inadvertent fuel addition.
3.1. Reactor core The fuel is in the form of flat plates of an alloy of 233 U and aluminum. These plates are assembled in an aluminum casing to form the fuel subassembly (Fig. 1). Nine fuel subassemblies are arranged in a square lattice. Modular design of reflectors in different sizes and shapes facilitate ease of assembly around beam tubes. These modules are assembled around the core in a cubical arrangement for obtaining a fully reflected configuration (Figs. 2 and 3). Adjustable reflector blocks (ARB) are provided for flexibility to alter the configuration in order to gain the required excess
Fig. 1. Fuel subassembly.
874
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880 Table 2 Reactor physics and operation parameters of KAMINI at nominal power
Fig. 2. Cross section of reactor tank indicating core configuration.
reactivity and for compensating the loss of reactivity due to fuel burn up, thus obviating the need to replace the fuel subassemblies at periodic intervals. Storage locations of ARB 2 and 3 are also shown in Fig. 2. Three plutonium aluminum subassemblies are also available for trimming core excess reactivity within 0.91 $. These subassemblies are also used for identification of failed fuel by substitution method. Safety control plates made of cadmium sandwiched in aluminum are provided for reactor power control and shutdown. The two safety control plates are located at the core reflector periphery on the eastern side, moving in two rectangular water filled cavities adjacent to the core. Safety control plates are provided with a stepper motor drive system with stainless steel wire rope arrangement for controlled raising and lowering. Fine control of reactor power is possible owing to positional accuracy of 0.1 mm of the stepper motor drive. Gravity drop mechanism is provided for rapid shut down (scram) of the reactor. A solenoid latch prevents accidental lifting of the safety control plates from the top of the tank, without interfering with the lowering operation. 3.2. Irradiation facilities There are three beam tubes, two irradiation locations at the core reflector boundary and one location adjacent to the core
Fig. 3. Core cage.
Reactor operation time (h) Fuel burn up (max) (MWd/t) Temperature at core outlet (steady state) (◦ C) Temperature rise during longest run (10 h) (◦ C)
1050 1200 38 10
Flux (n/cm2 /s) Core average Peak Pneumatic fast transfer system location South beam tube (n/cm2 /s) West beam tube North beam tube South thimble (peak) North thimble (peak)
5.3 × 1012 8.0 × 1012 2.3 × 1012 1.70 × 108 1.70 × 108 6.54 × 107 3.21 × 1010 3.94 × 1010
Reactivity coefficients Void coefficient ($/ml) Power coefficient ($/kW) Moderator temperature coefficient ($/◦ C) Burn up reactivity coefficient ($/MWd) Integral worth of the safety control plates ($)
−0.023 −0.008 −0.017 0.363 8
for carrying out irradiation experiments. Flux available in these locations is given in Table 2. The beam tubes are located one each on the north, south and western side of the reactor vault. The south and north beam tubes are for neutron radiography and the west beam tube is for radiation physics research. These beam tubes extend from the core reflector boundary to a length of about 2 m and consist of inner Zircaloy-2 and outer stainless steel sections. The diameter of each beam tube varies from 50 mm at the core reflector boundary to 250 mm at the outer extremity. The beam tubes are normally kept closed with vertically moving motor operated beam shutters made of lead to reduce direct radiation streaming from the core and are opened only during irradiation experiments. The ratio of length to diameter (L/D) for south and north beam tubes is about 160 and the aperture size is 220 mm × 70 mm. The south beam tube is used for radioactive objects and the north beam tube for non-radioactive objects. The radioactive fuel subassembly is lowered from the top of the cell through a radiography tube by a carriage and drive system into the radiography location in front of the south beam tube. A mechanism with a rotation facility that can index the object in front of the beam is also provided enabling neutron tomography of fuel subassembly and the pins. Both film and real time radiography are possible. Specially designed fixed lead shield around the radiography tube and the two mobile shields made of lead and borated paraffin, movable in perpendicular directions protect the personnel from the highly active irradiated fuel. These mobile shields also provide access to the pit in front of the south beam tube for installing the radiography equipment. At the outer end of the west beam tube, meant for radiation physics experiments, a pit is provided to facilitate installation of the experimental set up. The pit is normally filled with concrete shield blocks. A special arrangement is made by providing removable paraffin shield blocks to create a cavity of 20 l in front of the west beam tube for locating test specimen for irradiation experiments and radiography.
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
A pneumatically operated fast sample transfer system (PFTS) is provided for irradiation of samples to study short lived isotopes. Samples with a maximum weight of 1.7 g in special polypropylene containers (20 mm diameter and 30 mm length) can be shot into the irradiation site located adjacent to the core reflector boundary during reactor operation and retrieved immediately after irradiation. The system can be operated in auto mode with provision for setting the duration of irradiation. Several samples can be irradiated sequentially in this location without shutting down the reactor. The sample loading station is equipped with a fume hood for sample preparation and preirradiation treatment and a gamma counting system for assay of short lived nuclei. Two in-tank locations on either side of the west beam tube outside the reflector blocks facilitate irradiation of samples up to 50 ml. These locations provide adequate flux for irradiation due to peaking of thermal neutron flux in the BeO reflector. The samples contained in a 50 ml aluminum container (thimble) can be lowered into the irradiation location by means of motorized drive mechanisms from the top of the reactor tank. In addition, the water cavity above the reactor core can be used for irradiation of samples located in specially designed fixtures. 3.3. Reactor systems The reactor tank is an open cylindrical stainless steel tank with a flat bottom having a diameter of 2 m and height of 4.19 m. The core reflector assembly is located on an aluminum base plate, which is supported on six pillars fixed to the bottom of the reactor tank. The reactor tank also accommodates the safety control plates, piping of coolant system, online demineralizer and waste disposal system, fuel storage boxes, spent fuel guide structures, neutron detectors and process instruments namely, thermocouples and level probes (Figs. 4 and 5). Air purging is maintained at the top plenum to flush out the hydrogen generated due to radiolytic dissociation of water in the reactor tank. A stainless steel catcher vessel of 400 mm height is provided around the
875
Fig. 5. View of reactor tank through water shield.
reactor tank at the bottom. In case of leak in reactor tank, water collects in the catcher vessel and leak detectors in the catcher vessel alert the operator. Floor of the reactor vault is lined with stainless steel to avoid water seepage. Specially made double hexagonal interlocking concrete bricks are used in the biological shield surrounding the reactor tank. The concrete shield up to a height of 3 m has a high density (4800 kg/m3 ) and thereafter blocks of normal density concrete are used to have an overall shield height of about 4.2 m. The shielding design is such that effective biological shielding is provided from two sources namely, the core and irradiated fuel of high burn up in the radiography location. About 3 m of water column above the core in the reactor tank provides adequate shielding for the personnel at the access platform on top of the reactor. Core cooling is by natural convection of water in the reactor tank. Primary and secondary coolant circuits consisting of water to water and water to air heat exchangers and pumps are provided to maintain constant water temperature during prolonged operation at high power to minimize reactivity variation due to rise in inlet water temperature for special experimental purposes. The piping terminations inside the reactor tank are suitably located to obtain effective mixing of water. An online demineralizer
Fig. 4. Vertical section of reactor tank.
876
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
(DM) unit maintains the water quality (pH 5.5–7.5, conductivity <3 S/cm and chloride <0.4 ppm) to minimize corrosion and radioactivity level in water. 3.4. Storage and handling Twelve fuel storage boxes are provided on either side of the core reflector assembly for temporary storage of the fresh or spent fuel assembly. Provision also exists for storage of adjustable and top axial reflectors away from the core. An under water carriage facilitates movement of fuel or other light loads under water within the reactor tank prior to discharge from the reactor. Fuel and reflector handling operations are carried out under water in shutdown state from a maintenance trolley positioned over the reactor tank. Special telescopic tools are used for handling to suit the limited headroom available above the reactor. A special lead cask is used for discharging the spent fuel. An overhead crane is provided for handling special fixtures for disposal of spent fuel and other components over the reactor tank. 3.5. Instrumentation and control system Neutronic and process instruments are provided for monitoring various system parameters and for safety action. The neutronic instrumentation consists of pulse and current channels. Two pulse channels using boron coated counters monitor the flux during shutdown and startup and four current channels with uncompensated ion chambers are provided for power range control and safety. The reactor protection system is a hardwired system adopting redundancy, diversity and failsafe features. It consists of two channels, each consisting of one group of redundant scram parameters and the protection system actuates scram based on one out of two logic. Scram check facility is incorporated to ensure that reactor is started only if scram circuit is healthy. All reactor operations are carried out from a central control room panel in manual mode (Fig. 6). The controls are configured as a hybrid system consisting of hardwired systems and microprocessor based data acquisition system.
Fig. 6. Control room.
Clad failure is detected by an increase in the water activity using an online monitor having a sodium iodide detector. The coolant pumps and beam shutters are provided with normal power supply. The control panel is provided with emergency power supply from the diesel generator. Battery backup is provided for important indications in the control room required for ascertaining safe shutdown after power failure. 4. Commissioning Commissioning of the reactor was started in 1995 after installation of the reactor tank components, coolant system piping, instrumentation and control panel. Installation of reflectors was a meticulous operation to achieve a compact assembly of reflector modules around penetrations to prevent neutron leakage. Excellent water quality and clarity were maintained by the mixed bed demineralizer plant throughout the commissioning phase. Water systems were commissioned and the reactor tank was filled with demineralized water. After getting safety clearance for fuel loading, first criticality was achieved in October 1996. For first criticality, the neutron detectors were kept close to the core to improve response at low power. After ensuring satisfactory performance of safety related systems at low power (0.5 W), the detector location was finalized to achieve monitoring capability up to nominal power level. In order to ensure reproducibility and consistency of critical heights and core reactivity status, a core restraining aluminum structure namely, core cage was installed on top of the core during initial phase of operation. The core cage is a lightweight structure facilitating tight seating of fuel elements, allowing easy handling and sized to protect against lifting under buoyancy effect. The reactor power was raised to 5 kW and then to nominal power of 30 kW in September 1997 after completion of absolute power calibration to determine the actual neutronic power, augmentation of shielding around the beam tubes and carrying out low power physics experiments. 5. Operation philosophy KAMINI reactor is operated during normal working hours for irradiation experiments. Generally, the operation is restricted to 200 kWh per week but it can be stretched for special experiments. The estimated burn up loss is around 0.18 $/y and core excess reactivity adjustment is planned using axial reflector blocks and plutonium subassembly to within 0.91 $. The excess reactivity is provided for experimental load, temperature rise and only partially for xenon build up thereby restricting the continuous operating time to 12 h at 30 kW power. All operations and sample irradiations are carried out in accordance with well laid out operating procedures under strict administrative control and health physics surveillance (Technical Specifications for KAMINI operation, 1998). The performance of various systems are reviewed periodically and improvements carried out based on operational feedback. Since the reactor is operated only during normal working hours, for exigencies, signals like radiation level, fire, water
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
877
logging are monitored continuously in a separate control room, which is manned round the clock. 6. Operating experience 6.1. Physics experiments First criticality was achieved with nine fuel subassemblies and adjustable reflector blocks. Subassemblies were loaded progressively replacing the dummy graphite subassemblies and inverse count rate method was adopted to predict criticality. The absolute power calibration by thermal balance method could not be employed in this reactor since the core cooling is by natural convection and the coolant temperature changes are very small during reactor operation. Therefore, the power calibration was carried out by the irradiation of uranium–aluminum wires. The axial and radial neutron flux shapes in the core and the total fission power were estimated from the measured fission product activity of the wires. To reduce the gamma effect at detector location during nominal power operation, lead shields were installed in front of the detectors. Neutron flux at the detector location was found increased due to lower attenuation of lead shield compared to water, necessitating recalibration. Later on, gold foils were irradiated at the pneumatic fast transfer location to establish the correspondence of the gold foil activity with the reactor power. Subsequently for periodic verification of power calibration, gold foil activation method has been standardized. Neutron flux levels in the radial and the axial directions were mapped by gold foil irradiation. Typical axial flux distribution is shown in Fig. 7. The safety parameters of utmost importance in KAMINI are the worth of the safety control plates, moderator temperature coefficient of reactivity and void coefficient of reactivity. Changes in the core reactivity of KAMINI are usually estimated in terms of the changes in the position of the safety control plates. Hence calibration of the safety control plates (determination of reactivity worth per unit movement of safety control plate) is essential. Calibration of the worth was carried out in KAMINI before measuring the other safety parameters. The integral worth of both the safety control plates was measured again by rod drop method and found to be 8 ± 0.48 $
Fig. 8. Safety control plate worth shape.
(Mohapatra et al., 2004). The cumulative worth shape of individual safety control plate was obtained by sub-critical multiplication method. Fig. 8 depicts the experimentally measured and the fifth order polynomial worth shape of the safety control plate-2. The total worth of safety control plate-2 is 4.52 $ (Fig. 8) The void coefficient of reactivity in KAMINI was measured in different locations inside the core (refer Table 3) (Mohapatra, 2004). This was accomplished; by creating voids at different locations in the core and recording the subsequent changes in the reactivity. The fuel plates of KAMINI have 6 mm gap between them, which is filled by water. For measuring the void worth in a particular fuel subassembly location, special aluminum boxes were introduced between the fuel plates to create the void space. These boxes were attached to aluminum holding tubes, which were clamped to the reactor top. The reactor was made critical and was maintained at a low power and safety control plate-2 position was noted. The void box was withdrawn after shutting down the reactor. The void box was punctured and reintroduced in the same location so that the void is filled with water. The reactor was made critical and the position of safety control plate-2 was noted. The void worth was computed from the difference in critical heights as given below: ρvoid = ρair-box − ρwater-box where ρvoid is the negative reactivity introduced in the core. Measured values of void coefficient in various locations are given in Table 3. Measurement of moderator temperature coefficient of reactivity was performed by heating up the reactor tank water, which acts as moderator and coolant and compensation of reactivity loss was done by adjusting the safety control plate position. Eleven U-type electrical immersion heaters (33 kW power) were provided in the reactor tank to facilitate heating of water during Table 3 Void reactivity coefficients at different locations
Fig. 7. Axial flux distribution.
Location of void
Void coefficient ($/ml)
South west corner subassembly Central subassembly Subassembly adjacent to safety control plates North central subassembly
−0.01 ± 0.001 −0.02 ± 0.001 −0.02 ± 0.002 −0.008 ± 0.00042
878
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
Table 4 Moderator temperature coefficient measurement experimental results Inlet temperature (◦ C)
Safety control plate-2 position (mm)
Negative reactivity addition ($)
ρ due to temperature ($)
ρ/T ($/◦ C)
29 33 37 41 43
200.6 205.8 211.5 217.5 220.9
0 (reference) 0.081 0.042 0.062 0.072
– −0.07 −0.07 −0.07 −0.03
– −0.017 −0.017 −0.017 −0.017
Table 5 Measured neutron flux (reactor power 20 W) Beam tube
Thermal flux, n/cm2 /s (≤0.4 eV)
Flux, n/cm2 /s (up to 10 keV)
Total flux, n/cm2 /s
West North South
5.12 E 04 1.45 E 04 2.87 E 04
8.36 E 04 2.94 E 04 7.08 E 04
1.13 E 05 4.36 E 04 1.13 E 05
the measurement. The reactor was made critical with heaters on and power was maintained at 10 W. The primary coolant system was kept in service to ensure uniform heating of the tank water. The reactor power was maintained at 10 W by adjusting the safety control plate-2 position as the temperature of the tank water was rising gradually. The experiment was carried out for 7 h and the overall temperature rise was 14 ◦ C (29–43 ◦ C). The measured moderator temperature coefficient for different temperature ranges is given in Table 4. The average moderator temperature coefficient is found to be 0.017 ± 0.0014 $/◦ C (Mohapatra and Mohanakrishnan, 2000). The neutron flux measurements at the three beam tube locations were carried out by multi-foil activation method. The foils were irradiated with and without cadmium cover to determine the total and epithermal component of the flux (Table 5) (Mohapatra and Mohanakrishnan, 2002). 6.2. Fuel performance The 233 U fuel has achieved a burn up of about 1200 MWd/t till the end of December 2004. The burn up reactivity variation is in agreement with prediction (Table 6) as seen by the comparison of predicted and measured critical heights. Operation experience has indicated that the low delayed neutron fraction of the fuel does not pose problem for smooth power control. The xenon evolution is lower compared to 235 U based fuel resulting in no post-shutdown build up of xenon though the flux level is 5.3 × 1012 n/cm2 /s. This is because, 135 I yield which is precursor to 135 Xe is 4.9% for 233 U fission which is less compared to 6.2% for 235 U fission. Table 6 Comparison of expected and observed critical heights (reactor power 10 W) Year
Total fuel burn up (MWd/t)
Expected critical height of safety control plate-2 (mm)
Observed critical height of safety control plate-2 (mm)
1999–2000 2000–2001 2001–2002
365 546 750
200.2 201.8 218.9
200.1 201.5 219
Conduciveness of plate type fuel with water gap for natural convection flow and heat transfer has been established at all power levels. Fuel assembly design facilitates ease of handling. There are no apparent signs of corrosion in demineralized water medium during 7 years of operation indicating good compatibility of fuel with the medium. 6.3. Reactor system performance The reactor has been operated at various powers and durations for meeting the requirements of users. Quick startup of reactor in about 40 min is possible to meet the experimental requirements. Operation and physics parameters are given in Table 2. During reactor operation, one of the safety control plate-1 is kept fully withdrawn and the safety control plate-2 is used for reactor power adjustment. It has been demonstrated that the reactor can be operated at 30 kW for a maximum of 10 h for special experiments. This has provided valuable data on available excess reactivity, which is required for planning long duration experiments. Power fluctuations of 1–1.5 kW during steady power operation are easily controllable with the safety control plate adjustment. Xenon poisoning sets in after about 1 h of operation requiring compensation by safety control plate withdrawal. During an experimental 10 h run at 30 kW, the total reactivity compensation was 0.35 $, out of which 0.14 $ accounted for temperature feed back effect, the rest being for xenon compensation. Safety control plate withdrawal during the first 1 h only accounted for temperature rise. Subsequently the safety control plate withdrawal was more, indicating compensation for xenon poison effect and temperature. The temperature fluctuation of 7–8 ◦ C observed at core outlet indicates setting up of natural convection currents. Water quality and clarity in the tank has been well maintained by periodic circulation through the demineralizer unit preventing corrosion of structural components, radioactivity build up and providing clear view of the core for visual inspection and handling of in-tank components. Piping modifications were carried out inside the reactor tank to avoid passage of bubbles through the core during pump startup as they can introduce negative reactivity. The technical specification limit on water activity was modified taking into account activity due to the presence of short lived fission gases attributed to minor contamination of fuel clad surface. Handling of fuel and reflector needing careful judgment and dexterity in handling of tools under water have been carried out manually by maneuvering the tools from the top with out instrument backup. Experience at KAMINI has been incident free
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
879
Table 7 Radiation levels in different areas of KAMINI (reactor power 30 kW) Area
β, γ (Sv/h)
Neutron (Sv/h)
Radiation physics area Operating area at radio metallurgy laboratories Activation analysis lab
0.3 Background
Background Background
0.3
Background
Acceptable limit 1.0 Sv/h.
and smooth, affirming confidence for carrying out such critical operations safely. Performance of instrumentation and control systems is satisfactory. To prevent neutron detector failure due to moisture ingress, mineral insulated cables were replaced by moisture resistant polyethylene cables. Noise pickup problems were solved by replacing electronic components and rerouting of cables. Log P trip setting was raised from 110% to 130% to increase the operating margin. Spurious trips on log P from one of the redundant neutronic channels due to fleeting loose contact in the relay base were eliminated by providing clamps to secure the relay to its base to ensure tight contact. Uncertainty regarding smooth transit of sample in the pneumatic fast transfer system sample injection tube were experienced in the early stages. Correction of minor eccentricity in the piping and modification of the sample capsule geometry to cylindrical shape with chamfered edges resulted in smooth injection and ejection of the sample capsule. After carrying out irradiation of large number and different types of samples at various power levels and duration, the control station was relocated away from the loading station for the safety of the operator from exposure to activity of samples. All sample handling and irradiations have been carried out without contamination and exposure to plant personnel. Radiation levels at various locations are within permissible limits (Table 7). 7. Utilization for R&D 7.1. Neutron radiography of fuel South beam tube has been used extensively for neutron radiography because of the good length to diameter ratio (L/D), mobile shielding provision and ease of access to radiography area. In the first part of the campaign, characterization of neutron beam was carried out. The objects examined were the beam purity indicator, sensitivity indicator (fuel). Both direct and indirect imaging techniques were used. In the next phase dummy fuel pins of fast reactor and pressurized heavy water reactor were radiographed. Transfer technique was used for imaging of these fuel pins. A number of experiments were conducted at various power levels and exposure time of 30 min. Resolution of images was found to be satisfactory and pellet to pellet gaps and chipped pellets could be detected. Neutron radiography of fast reactor fuel pins of 25, 50
Fig. 9. Cable cutter and its neutron radiographic image.
and 100 GWd/t burn up was carried out for characterization of fuel and pellet to pellet gap (Ramanarayanan et al., 2002). 7.2. Neutron radiography of Pyro devices Pyro devices are extensively used in the space industry. These are basically mechanical devices with a small amount of explosive. Different types of these devices are used for ignition, shearing the straps, cables and bolt cutters (Fig. 9). Ensuring the reliability of the Pyro is a very crucial aspect in the space program. Of all the nondestructive examination (NDE) techniques, the neutron radiography is the best for this purpose. KAMINI is being used extensively for the qualification of Pyros for space missions of the Indian Space Research Organization (ISRO). About 1600 such components have been radiographed in eleven campaigns (Ramanarayanan et al., 2002). 7.3. Neutron activation analysis Irradiation of various samples has been carried out at pneumatic fast transfer system and thimble locations for activation analysis. Neutron activation analysis is a sensitive nondestructive and multi-elemental technique. About 375 samples have been irradiated in KAMINI. Typical applications of neutron activation analysis using KAMINI include analysis of geological samples like ores, rocks and chemical samples from the forensic laboratories and development of method for neptunium (237 Np) estimation at microgram levels for use in reprocessing industry. Ore samples were assayed for gold and platinum group elements. Assay of iodine in leaf samples as well as rock samples for rare earths and other elements were carried out for environmental studies. Fire retardant paint of cables was analyzed for bromine content. 8. Conclusion KAMINI is a unique 233 U fuelled neutron source facility operating in India. It is providing R&D facilities for neutron
880
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
radiography, activation analysis and radiation physics experiments. Seven years of operating experience with this facility has been satisfactory. Efforts are underway to enhance utilization of KAMINI for radiation physics and shielding experiments by augmentation of facilities for irradiation. References Dietrich, J.P., 1958. Experimental determination of self regulation and safety of operating water moderated reactor. Geneva Conference Paper. Mohapatra, D.K., 2004. Experimental and theoretical investigations in a thermal neutron research reactor – KAMINI. Doctoral Thesis. Submitted to the University of Madras, India. Mohapatra, D.K., Mohanakrishnan, P., 2000. Moderator temperature effect on reactivity in light water moderated experimental reactors. Ann. Nucl. Energy 27, 969–983. Mohapatra, D.K., Mohanakrishnan, P., 2002. Measurement and prediction of neutron spectra in the Kalpakkam Mini Reactor (KAMINI). Appl. Radiat. Isotopes 57, 25. Mohapatra, D.K., Radha, E., Mohanakrishnan, P., 2004. Theoretical and experimental investigations of reactor parameters in a
U-233 fuelled research reactor. Ann. Nucl. Energy 31, 197– 212. Pasupathy, C.S., et al., 1990. Physics and instrumentation of 233 U fuelled neutron source reactor KAMINI, BARC-1532, p. 160. Radkowsky, A., 1986. Seed and Blanket Reactors, CRC Handbook on Nuclear Reactor Calculations, III. CRC Press, Florida, USA. Ramanarayanan, R.R., Anandkumar, V., Mohanakrishnan, P., Pillai, C.P., Kumar, P.V., Kapoor, R.P., 1999. KAMINI reactor commissioning and operating experience, research facilities and their utilization. In: Proceedings of the IAEA Symposium on Research Reactor Utilization, Safety and Management, Portugal. Ramanarayanan, R.R., Venkataraman, B., Raghu, N., Johny, T., Ramalingam, P.V., Kapoor, R.P., 2002. Utilization of KAMINI reactor for neutron radiography and activation analysis. In: Proceedings of the National Seminar on NDE (NDE 2002), Chennai, December 5–7, 2002. Singh, K., et al., 1980. Adequacy of a time constant concept for the estimation of fuel assembly temperatures in nuclear reactor power transients. In: Paper Presented at the 5th National Heat and Mass Transfer Conference, Hyderabad, February. Technical Specifications for KAMINI operation, 1998. Atomic Energy Regulatory Board, India.
Research reactor KAMINI S. Usha ∗ , R.R. Ramanarayanan ∗ , P. Mohanakrishnan, R.P. Kapoor Indira Gandhi Centre for Atomic Research, Kalpakkam, India Received 11 March 2004; received in revised form 26 September 2005; accepted 29 September 2005
Abstract KAMINI is a 30 kW, 233 U fuelled research reactor located at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. The reactor functions as a neutron source with a flux of 1012 n/cm2 /s at the core center. KAMINI uses plate type of fuel in a reactor tank. Demineralized water is used as moderator, biological shield and coolant. Reflectors are made of beryllium oxide with Zircaloy-2 sheath. Cadmium is used as the absorber material in the safety control plates (SCP) provided for power control and shutdown of the reactor. There are three beam tubes and three irradiation locations within the reactor tank facilitating neutron radiography of radioactive and nonradioactive objects, sample irradiation for neutron activation analysis, physics and shielding experiments. The reactor attained first criticality in October 1996 and nominal power of 30 kW in September 1997. This article covers the design description, facilities available for experiments and their utilization for research and development. Seven years of operating experience of KAMINI, which is the only reactor using 233 U as fuel, are also highlighted. © 2006 Elsevier B.V. All rights reserved.
1. Introduction
2. Basic design features
KAMINI (KAlpakkam MINI) is a 30 kW, 233 U fuelled, demineralized light water moderated and cooled, special purpose research reactor located at Indira Gandhi Centre for Atomic Research (IGCAR). Beryllium oxide (BeO) is used as reflector and cadmium is used as absorber material in the safety control plates (233 U fuelled reactors have been operated earlier in many countries, notably, the Shippingport light water breeder reactor in US). The reactor was designed and built jointly by Bhabha Atomic Research Centre (BARC) and IGCAR. Salient features of KAMINI are given in Table 1. The reactor functions as a neutron source with a flux of 1012 n/cm2 /s at the core center and facilitates carrying out neutron radiography of both radioactive and non-radioactive objects and neutron activation analysis. Facilities are also available in the reactor for carrying out radiation physics research, irradiation of large samples and calibration and testing of neutron detectors (Ramanarayanan et al., 1999).
A judicious choice of design features, layout and control philosophy has been made using the experimental results and experience of similar reactors to provide operational flexibility and to achieve inherent safety. That is, in case of any power excursion due to inadvertent addition of reactivity, the reactor shuts down due to negative void and temperature coefficient. KAMINI is a tank type reactor using enriched uranium and aluminum plate type fuel in a light water pool. Physics of 233 U fuelled system including a full scale mock up of KAMINI core has been studied extensively in PURNIMA reactor at BARC to establish the neutronic parameters and safety (Pasupathy et al., 1990; Radkowsky, 1986; Dietrich, 1958). KAMINI is a reflector moderated reactor where 50% of fissions are due to reflector returned neutrons. Zircoloy-2 canned BeO is used as reflector owing to its high reflection efficiency resulting in lower fuel inventory. Choice of Zircaloy-2 sheath for BeO is due to its superiority over aluminum with respect to neutron economy and structural considerations. Fully reflected reactor concept (reflectors are arranged on all sides around the core) is adopted to minimize fuel inventory and to prevent inadvertent addition of large positive reactivity by reflector movement.
∗
Corresponding authors. E-mail addresses: [email protected] (S. Usha), [email protected] (R.R. Ramanarayanan). 0029-5493/$ – see front matter © 2006 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2005.09.033
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880 Table 1 Salient characteristics of KAMINI Reactor type Nominal power (kW) Core size (mm) Fuel Fuel inventory (g) Subassemblies Plates per subassembly Gap between fuel plates (mm) Plate size (mm) Subassembly size (mm) Reflector Number of modules Number of adjustable reflectors Absorber material Excess reactivity ($) Effective delayed neutron fraction ($) Reactivity compensation Moderator/coolant/shield Quantity of moderator (t)
Open tank type (2 m diameter, 4.19 m height) 30 204 × 204 × 275 233 U
(20%)–Al alloy plates with Al clad 600 9 8 6 2 × 62 × 260 66 × 66 × 275 BeO (200 mm thick) encased in Zircaloy-2 sheath 20 3 Cadmium sandwiched in aluminium 0.76 1 (βeff = 0.0033) By reflector addition Demineralized water (pH: 5.5–7.5, K ≤3 S/cm) 12.5
873
Possibility of reactivity addition by the sample, through the pneumatic fast transfer system used for study of short lived activity is eliminated by the restriction on sample volume and strict administrative controls. The reactor core is cooled by natural convection. Any sudden power increase (as in an accident) results in a rapid increase of temperature in the vicinity of the core. This causes shutdown of the reactor due to high negative void and temperature coefficient of reactivity. The reactor vault is located in the basement of the hot cell, located at the ground level to enable lowering of irradiated fuel in front of the beam tube directly from the operating floor of the hot cell. Floor of the reactor vault is 1.5 m below the floor level of the basement area and the core top is 300 mm below the basement floor level with 3 m of water column above the core in the reactor tank, ensuring submergence of the core in case of gross leak in the reactor tank. The building housing the reactor is located at a flood safe elevation and double isolation valves are provided on the demineralized water supply line from outside to prevent accidental flooding in the basement. Stainless steel is used as the material for the reactor tank and coolant system piping with welded joints to minimize the possibility of leak. 3. Design description
The inherent safety of KAMINI reactor arises from the negative temperature coefficient (−0.017 $/◦ C) and the negative void coefficient (−0.0230 $/ml), which act as shutdown mechanisms in case of insertion of excess reactivity. A small inventory of 233 U (approximately 0.6 kg) used in a compact core of about 10 l limits the radiological hazard potential during accidental conditions. As indicated in the safety analysis, for uncontrolled withdrawal of the safety control plate, the maximum power reached is only 115 kW, which can be easily removed by the reactor tank water with the water temperature not exceeding 70 ◦ C without any safety consequences (the reactor safety was evaluated using SECMOD code) (Singh et al., 1980). Physics design with the H/233 U ratio of around 320, being in the under moderated region, ensures that moderator temperature coefficient is negative providing operational stability. Reactivity insertion rate limited to 0.009 $/mm allows enough time for preventive action in case of loss of regulation. Total worth of absorber plates is such that adequate shutdown margin (more than 3.030 $ with one safety control plate withdrawn) is provided while allowing a reasonably rapid startup for experimental purposes. The excess reactivity is provided to fully compensate for the essential operating loads namely, temperature and sample in the irradiation locations but only partial compensation is provided for xenon poisoning. This limits the excess reactivity and acts as self shutdown mechanism if the reactor is operated for more than 10 h at 30 kW power. Excess reactivity of less than 1 $ (βeff = 0.0033) ensures that the reactor cannot attain prompt criticality. Reactivity compensation for burn up loss by reflector addition minimizes handling of fuel. There are no vacant locations in the core, which prevents inadvertent fuel addition.
3.1. Reactor core The fuel is in the form of flat plates of an alloy of 233 U and aluminum. These plates are assembled in an aluminum casing to form the fuel subassembly (Fig. 1). Nine fuel subassemblies are arranged in a square lattice. Modular design of reflectors in different sizes and shapes facilitate ease of assembly around beam tubes. These modules are assembled around the core in a cubical arrangement for obtaining a fully reflected configuration (Figs. 2 and 3). Adjustable reflector blocks (ARB) are provided for flexibility to alter the configuration in order to gain the required excess
Fig. 1. Fuel subassembly.
874
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880 Table 2 Reactor physics and operation parameters of KAMINI at nominal power
Fig. 2. Cross section of reactor tank indicating core configuration.
reactivity and for compensating the loss of reactivity due to fuel burn up, thus obviating the need to replace the fuel subassemblies at periodic intervals. Storage locations of ARB 2 and 3 are also shown in Fig. 2. Three plutonium aluminum subassemblies are also available for trimming core excess reactivity within 0.91 $. These subassemblies are also used for identification of failed fuel by substitution method. Safety control plates made of cadmium sandwiched in aluminum are provided for reactor power control and shutdown. The two safety control plates are located at the core reflector periphery on the eastern side, moving in two rectangular water filled cavities adjacent to the core. Safety control plates are provided with a stepper motor drive system with stainless steel wire rope arrangement for controlled raising and lowering. Fine control of reactor power is possible owing to positional accuracy of 0.1 mm of the stepper motor drive. Gravity drop mechanism is provided for rapid shut down (scram) of the reactor. A solenoid latch prevents accidental lifting of the safety control plates from the top of the tank, without interfering with the lowering operation. 3.2. Irradiation facilities There are three beam tubes, two irradiation locations at the core reflector boundary and one location adjacent to the core
Fig. 3. Core cage.
Reactor operation time (h) Fuel burn up (max) (MWd/t) Temperature at core outlet (steady state) (◦ C) Temperature rise during longest run (10 h) (◦ C)
1050 1200 38 10
Flux (n/cm2 /s) Core average Peak Pneumatic fast transfer system location South beam tube (n/cm2 /s) West beam tube North beam tube South thimble (peak) North thimble (peak)
5.3 × 1012 8.0 × 1012 2.3 × 1012 1.70 × 108 1.70 × 108 6.54 × 107 3.21 × 1010 3.94 × 1010
Reactivity coefficients Void coefficient ($/ml) Power coefficient ($/kW) Moderator temperature coefficient ($/◦ C) Burn up reactivity coefficient ($/MWd) Integral worth of the safety control plates ($)
−0.023 −0.008 −0.017 0.363 8
for carrying out irradiation experiments. Flux available in these locations is given in Table 2. The beam tubes are located one each on the north, south and western side of the reactor vault. The south and north beam tubes are for neutron radiography and the west beam tube is for radiation physics research. These beam tubes extend from the core reflector boundary to a length of about 2 m and consist of inner Zircaloy-2 and outer stainless steel sections. The diameter of each beam tube varies from 50 mm at the core reflector boundary to 250 mm at the outer extremity. The beam tubes are normally kept closed with vertically moving motor operated beam shutters made of lead to reduce direct radiation streaming from the core and are opened only during irradiation experiments. The ratio of length to diameter (L/D) for south and north beam tubes is about 160 and the aperture size is 220 mm × 70 mm. The south beam tube is used for radioactive objects and the north beam tube for non-radioactive objects. The radioactive fuel subassembly is lowered from the top of the cell through a radiography tube by a carriage and drive system into the radiography location in front of the south beam tube. A mechanism with a rotation facility that can index the object in front of the beam is also provided enabling neutron tomography of fuel subassembly and the pins. Both film and real time radiography are possible. Specially designed fixed lead shield around the radiography tube and the two mobile shields made of lead and borated paraffin, movable in perpendicular directions protect the personnel from the highly active irradiated fuel. These mobile shields also provide access to the pit in front of the south beam tube for installing the radiography equipment. At the outer end of the west beam tube, meant for radiation physics experiments, a pit is provided to facilitate installation of the experimental set up. The pit is normally filled with concrete shield blocks. A special arrangement is made by providing removable paraffin shield blocks to create a cavity of 20 l in front of the west beam tube for locating test specimen for irradiation experiments and radiography.
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
A pneumatically operated fast sample transfer system (PFTS) is provided for irradiation of samples to study short lived isotopes. Samples with a maximum weight of 1.7 g in special polypropylene containers (20 mm diameter and 30 mm length) can be shot into the irradiation site located adjacent to the core reflector boundary during reactor operation and retrieved immediately after irradiation. The system can be operated in auto mode with provision for setting the duration of irradiation. Several samples can be irradiated sequentially in this location without shutting down the reactor. The sample loading station is equipped with a fume hood for sample preparation and preirradiation treatment and a gamma counting system for assay of short lived nuclei. Two in-tank locations on either side of the west beam tube outside the reflector blocks facilitate irradiation of samples up to 50 ml. These locations provide adequate flux for irradiation due to peaking of thermal neutron flux in the BeO reflector. The samples contained in a 50 ml aluminum container (thimble) can be lowered into the irradiation location by means of motorized drive mechanisms from the top of the reactor tank. In addition, the water cavity above the reactor core can be used for irradiation of samples located in specially designed fixtures. 3.3. Reactor systems The reactor tank is an open cylindrical stainless steel tank with a flat bottom having a diameter of 2 m and height of 4.19 m. The core reflector assembly is located on an aluminum base plate, which is supported on six pillars fixed to the bottom of the reactor tank. The reactor tank also accommodates the safety control plates, piping of coolant system, online demineralizer and waste disposal system, fuel storage boxes, spent fuel guide structures, neutron detectors and process instruments namely, thermocouples and level probes (Figs. 4 and 5). Air purging is maintained at the top plenum to flush out the hydrogen generated due to radiolytic dissociation of water in the reactor tank. A stainless steel catcher vessel of 400 mm height is provided around the
875
Fig. 5. View of reactor tank through water shield.
reactor tank at the bottom. In case of leak in reactor tank, water collects in the catcher vessel and leak detectors in the catcher vessel alert the operator. Floor of the reactor vault is lined with stainless steel to avoid water seepage. Specially made double hexagonal interlocking concrete bricks are used in the biological shield surrounding the reactor tank. The concrete shield up to a height of 3 m has a high density (4800 kg/m3 ) and thereafter blocks of normal density concrete are used to have an overall shield height of about 4.2 m. The shielding design is such that effective biological shielding is provided from two sources namely, the core and irradiated fuel of high burn up in the radiography location. About 3 m of water column above the core in the reactor tank provides adequate shielding for the personnel at the access platform on top of the reactor. Core cooling is by natural convection of water in the reactor tank. Primary and secondary coolant circuits consisting of water to water and water to air heat exchangers and pumps are provided to maintain constant water temperature during prolonged operation at high power to minimize reactivity variation due to rise in inlet water temperature for special experimental purposes. The piping terminations inside the reactor tank are suitably located to obtain effective mixing of water. An online demineralizer
Fig. 4. Vertical section of reactor tank.
876
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
(DM) unit maintains the water quality (pH 5.5–7.5, conductivity <3 S/cm and chloride <0.4 ppm) to minimize corrosion and radioactivity level in water. 3.4. Storage and handling Twelve fuel storage boxes are provided on either side of the core reflector assembly for temporary storage of the fresh or spent fuel assembly. Provision also exists for storage of adjustable and top axial reflectors away from the core. An under water carriage facilitates movement of fuel or other light loads under water within the reactor tank prior to discharge from the reactor. Fuel and reflector handling operations are carried out under water in shutdown state from a maintenance trolley positioned over the reactor tank. Special telescopic tools are used for handling to suit the limited headroom available above the reactor. A special lead cask is used for discharging the spent fuel. An overhead crane is provided for handling special fixtures for disposal of spent fuel and other components over the reactor tank. 3.5. Instrumentation and control system Neutronic and process instruments are provided for monitoring various system parameters and for safety action. The neutronic instrumentation consists of pulse and current channels. Two pulse channels using boron coated counters monitor the flux during shutdown and startup and four current channels with uncompensated ion chambers are provided for power range control and safety. The reactor protection system is a hardwired system adopting redundancy, diversity and failsafe features. It consists of two channels, each consisting of one group of redundant scram parameters and the protection system actuates scram based on one out of two logic. Scram check facility is incorporated to ensure that reactor is started only if scram circuit is healthy. All reactor operations are carried out from a central control room panel in manual mode (Fig. 6). The controls are configured as a hybrid system consisting of hardwired systems and microprocessor based data acquisition system.
Fig. 6. Control room.
Clad failure is detected by an increase in the water activity using an online monitor having a sodium iodide detector. The coolant pumps and beam shutters are provided with normal power supply. The control panel is provided with emergency power supply from the diesel generator. Battery backup is provided for important indications in the control room required for ascertaining safe shutdown after power failure. 4. Commissioning Commissioning of the reactor was started in 1995 after installation of the reactor tank components, coolant system piping, instrumentation and control panel. Installation of reflectors was a meticulous operation to achieve a compact assembly of reflector modules around penetrations to prevent neutron leakage. Excellent water quality and clarity were maintained by the mixed bed demineralizer plant throughout the commissioning phase. Water systems were commissioned and the reactor tank was filled with demineralized water. After getting safety clearance for fuel loading, first criticality was achieved in October 1996. For first criticality, the neutron detectors were kept close to the core to improve response at low power. After ensuring satisfactory performance of safety related systems at low power (0.5 W), the detector location was finalized to achieve monitoring capability up to nominal power level. In order to ensure reproducibility and consistency of critical heights and core reactivity status, a core restraining aluminum structure namely, core cage was installed on top of the core during initial phase of operation. The core cage is a lightweight structure facilitating tight seating of fuel elements, allowing easy handling and sized to protect against lifting under buoyancy effect. The reactor power was raised to 5 kW and then to nominal power of 30 kW in September 1997 after completion of absolute power calibration to determine the actual neutronic power, augmentation of shielding around the beam tubes and carrying out low power physics experiments. 5. Operation philosophy KAMINI reactor is operated during normal working hours for irradiation experiments. Generally, the operation is restricted to 200 kWh per week but it can be stretched for special experiments. The estimated burn up loss is around 0.18 $/y and core excess reactivity adjustment is planned using axial reflector blocks and plutonium subassembly to within 0.91 $. The excess reactivity is provided for experimental load, temperature rise and only partially for xenon build up thereby restricting the continuous operating time to 12 h at 30 kW power. All operations and sample irradiations are carried out in accordance with well laid out operating procedures under strict administrative control and health physics surveillance (Technical Specifications for KAMINI operation, 1998). The performance of various systems are reviewed periodically and improvements carried out based on operational feedback. Since the reactor is operated only during normal working hours, for exigencies, signals like radiation level, fire, water
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
877
logging are monitored continuously in a separate control room, which is manned round the clock. 6. Operating experience 6.1. Physics experiments First criticality was achieved with nine fuel subassemblies and adjustable reflector blocks. Subassemblies were loaded progressively replacing the dummy graphite subassemblies and inverse count rate method was adopted to predict criticality. The absolute power calibration by thermal balance method could not be employed in this reactor since the core cooling is by natural convection and the coolant temperature changes are very small during reactor operation. Therefore, the power calibration was carried out by the irradiation of uranium–aluminum wires. The axial and radial neutron flux shapes in the core and the total fission power were estimated from the measured fission product activity of the wires. To reduce the gamma effect at detector location during nominal power operation, lead shields were installed in front of the detectors. Neutron flux at the detector location was found increased due to lower attenuation of lead shield compared to water, necessitating recalibration. Later on, gold foils were irradiated at the pneumatic fast transfer location to establish the correspondence of the gold foil activity with the reactor power. Subsequently for periodic verification of power calibration, gold foil activation method has been standardized. Neutron flux levels in the radial and the axial directions were mapped by gold foil irradiation. Typical axial flux distribution is shown in Fig. 7. The safety parameters of utmost importance in KAMINI are the worth of the safety control plates, moderator temperature coefficient of reactivity and void coefficient of reactivity. Changes in the core reactivity of KAMINI are usually estimated in terms of the changes in the position of the safety control plates. Hence calibration of the safety control plates (determination of reactivity worth per unit movement of safety control plate) is essential. Calibration of the worth was carried out in KAMINI before measuring the other safety parameters. The integral worth of both the safety control plates was measured again by rod drop method and found to be 8 ± 0.48 $
Fig. 8. Safety control plate worth shape.
(Mohapatra et al., 2004). The cumulative worth shape of individual safety control plate was obtained by sub-critical multiplication method. Fig. 8 depicts the experimentally measured and the fifth order polynomial worth shape of the safety control plate-2. The total worth of safety control plate-2 is 4.52 $ (Fig. 8) The void coefficient of reactivity in KAMINI was measured in different locations inside the core (refer Table 3) (Mohapatra, 2004). This was accomplished; by creating voids at different locations in the core and recording the subsequent changes in the reactivity. The fuel plates of KAMINI have 6 mm gap between them, which is filled by water. For measuring the void worth in a particular fuel subassembly location, special aluminum boxes were introduced between the fuel plates to create the void space. These boxes were attached to aluminum holding tubes, which were clamped to the reactor top. The reactor was made critical and was maintained at a low power and safety control plate-2 position was noted. The void box was withdrawn after shutting down the reactor. The void box was punctured and reintroduced in the same location so that the void is filled with water. The reactor was made critical and the position of safety control plate-2 was noted. The void worth was computed from the difference in critical heights as given below: ρvoid = ρair-box − ρwater-box where ρvoid is the negative reactivity introduced in the core. Measured values of void coefficient in various locations are given in Table 3. Measurement of moderator temperature coefficient of reactivity was performed by heating up the reactor tank water, which acts as moderator and coolant and compensation of reactivity loss was done by adjusting the safety control plate position. Eleven U-type electrical immersion heaters (33 kW power) were provided in the reactor tank to facilitate heating of water during Table 3 Void reactivity coefficients at different locations
Fig. 7. Axial flux distribution.
Location of void
Void coefficient ($/ml)
South west corner subassembly Central subassembly Subassembly adjacent to safety control plates North central subassembly
−0.01 ± 0.001 −0.02 ± 0.001 −0.02 ± 0.002 −0.008 ± 0.00042
878
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
Table 4 Moderator temperature coefficient measurement experimental results Inlet temperature (◦ C)
Safety control plate-2 position (mm)
Negative reactivity addition ($)
ρ due to temperature ($)
ρ/T ($/◦ C)
29 33 37 41 43
200.6 205.8 211.5 217.5 220.9
0 (reference) 0.081 0.042 0.062 0.072
– −0.07 −0.07 −0.07 −0.03
– −0.017 −0.017 −0.017 −0.017
Table 5 Measured neutron flux (reactor power 20 W) Beam tube
Thermal flux, n/cm2 /s (≤0.4 eV)
Flux, n/cm2 /s (up to 10 keV)
Total flux, n/cm2 /s
West North South
5.12 E 04 1.45 E 04 2.87 E 04
8.36 E 04 2.94 E 04 7.08 E 04
1.13 E 05 4.36 E 04 1.13 E 05
the measurement. The reactor was made critical with heaters on and power was maintained at 10 W. The primary coolant system was kept in service to ensure uniform heating of the tank water. The reactor power was maintained at 10 W by adjusting the safety control plate-2 position as the temperature of the tank water was rising gradually. The experiment was carried out for 7 h and the overall temperature rise was 14 ◦ C (29–43 ◦ C). The measured moderator temperature coefficient for different temperature ranges is given in Table 4. The average moderator temperature coefficient is found to be 0.017 ± 0.0014 $/◦ C (Mohapatra and Mohanakrishnan, 2000). The neutron flux measurements at the three beam tube locations were carried out by multi-foil activation method. The foils were irradiated with and without cadmium cover to determine the total and epithermal component of the flux (Table 5) (Mohapatra and Mohanakrishnan, 2002). 6.2. Fuel performance The 233 U fuel has achieved a burn up of about 1200 MWd/t till the end of December 2004. The burn up reactivity variation is in agreement with prediction (Table 6) as seen by the comparison of predicted and measured critical heights. Operation experience has indicated that the low delayed neutron fraction of the fuel does not pose problem for smooth power control. The xenon evolution is lower compared to 235 U based fuel resulting in no post-shutdown build up of xenon though the flux level is 5.3 × 1012 n/cm2 /s. This is because, 135 I yield which is precursor to 135 Xe is 4.9% for 233 U fission which is less compared to 6.2% for 235 U fission. Table 6 Comparison of expected and observed critical heights (reactor power 10 W) Year
Total fuel burn up (MWd/t)
Expected critical height of safety control plate-2 (mm)
Observed critical height of safety control plate-2 (mm)
1999–2000 2000–2001 2001–2002
365 546 750
200.2 201.8 218.9
200.1 201.5 219
Conduciveness of plate type fuel with water gap for natural convection flow and heat transfer has been established at all power levels. Fuel assembly design facilitates ease of handling. There are no apparent signs of corrosion in demineralized water medium during 7 years of operation indicating good compatibility of fuel with the medium. 6.3. Reactor system performance The reactor has been operated at various powers and durations for meeting the requirements of users. Quick startup of reactor in about 40 min is possible to meet the experimental requirements. Operation and physics parameters are given in Table 2. During reactor operation, one of the safety control plate-1 is kept fully withdrawn and the safety control plate-2 is used for reactor power adjustment. It has been demonstrated that the reactor can be operated at 30 kW for a maximum of 10 h for special experiments. This has provided valuable data on available excess reactivity, which is required for planning long duration experiments. Power fluctuations of 1–1.5 kW during steady power operation are easily controllable with the safety control plate adjustment. Xenon poisoning sets in after about 1 h of operation requiring compensation by safety control plate withdrawal. During an experimental 10 h run at 30 kW, the total reactivity compensation was 0.35 $, out of which 0.14 $ accounted for temperature feed back effect, the rest being for xenon compensation. Safety control plate withdrawal during the first 1 h only accounted for temperature rise. Subsequently the safety control plate withdrawal was more, indicating compensation for xenon poison effect and temperature. The temperature fluctuation of 7–8 ◦ C observed at core outlet indicates setting up of natural convection currents. Water quality and clarity in the tank has been well maintained by periodic circulation through the demineralizer unit preventing corrosion of structural components, radioactivity build up and providing clear view of the core for visual inspection and handling of in-tank components. Piping modifications were carried out inside the reactor tank to avoid passage of bubbles through the core during pump startup as they can introduce negative reactivity. The technical specification limit on water activity was modified taking into account activity due to the presence of short lived fission gases attributed to minor contamination of fuel clad surface. Handling of fuel and reflector needing careful judgment and dexterity in handling of tools under water have been carried out manually by maneuvering the tools from the top with out instrument backup. Experience at KAMINI has been incident free
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
879
Table 7 Radiation levels in different areas of KAMINI (reactor power 30 kW) Area
β, γ (Sv/h)
Neutron (Sv/h)
Radiation physics area Operating area at radio metallurgy laboratories Activation analysis lab
0.3 Background
Background Background
0.3
Background
Acceptable limit 1.0 Sv/h.
and smooth, affirming confidence for carrying out such critical operations safely. Performance of instrumentation and control systems is satisfactory. To prevent neutron detector failure due to moisture ingress, mineral insulated cables were replaced by moisture resistant polyethylene cables. Noise pickup problems were solved by replacing electronic components and rerouting of cables. Log P trip setting was raised from 110% to 130% to increase the operating margin. Spurious trips on log P from one of the redundant neutronic channels due to fleeting loose contact in the relay base were eliminated by providing clamps to secure the relay to its base to ensure tight contact. Uncertainty regarding smooth transit of sample in the pneumatic fast transfer system sample injection tube were experienced in the early stages. Correction of minor eccentricity in the piping and modification of the sample capsule geometry to cylindrical shape with chamfered edges resulted in smooth injection and ejection of the sample capsule. After carrying out irradiation of large number and different types of samples at various power levels and duration, the control station was relocated away from the loading station for the safety of the operator from exposure to activity of samples. All sample handling and irradiations have been carried out without contamination and exposure to plant personnel. Radiation levels at various locations are within permissible limits (Table 7). 7. Utilization for R&D 7.1. Neutron radiography of fuel South beam tube has been used extensively for neutron radiography because of the good length to diameter ratio (L/D), mobile shielding provision and ease of access to radiography area. In the first part of the campaign, characterization of neutron beam was carried out. The objects examined were the beam purity indicator, sensitivity indicator (fuel). Both direct and indirect imaging techniques were used. In the next phase dummy fuel pins of fast reactor and pressurized heavy water reactor were radiographed. Transfer technique was used for imaging of these fuel pins. A number of experiments were conducted at various power levels and exposure time of 30 min. Resolution of images was found to be satisfactory and pellet to pellet gaps and chipped pellets could be detected. Neutron radiography of fast reactor fuel pins of 25, 50
Fig. 9. Cable cutter and its neutron radiographic image.
and 100 GWd/t burn up was carried out for characterization of fuel and pellet to pellet gap (Ramanarayanan et al., 2002). 7.2. Neutron radiography of Pyro devices Pyro devices are extensively used in the space industry. These are basically mechanical devices with a small amount of explosive. Different types of these devices are used for ignition, shearing the straps, cables and bolt cutters (Fig. 9). Ensuring the reliability of the Pyro is a very crucial aspect in the space program. Of all the nondestructive examination (NDE) techniques, the neutron radiography is the best for this purpose. KAMINI is being used extensively for the qualification of Pyros for space missions of the Indian Space Research Organization (ISRO). About 1600 such components have been radiographed in eleven campaigns (Ramanarayanan et al., 2002). 7.3. Neutron activation analysis Irradiation of various samples has been carried out at pneumatic fast transfer system and thimble locations for activation analysis. Neutron activation analysis is a sensitive nondestructive and multi-elemental technique. About 375 samples have been irradiated in KAMINI. Typical applications of neutron activation analysis using KAMINI include analysis of geological samples like ores, rocks and chemical samples from the forensic laboratories and development of method for neptunium (237 Np) estimation at microgram levels for use in reprocessing industry. Ore samples were assayed for gold and platinum group elements. Assay of iodine in leaf samples as well as rock samples for rare earths and other elements were carried out for environmental studies. Fire retardant paint of cables was analyzed for bromine content. 8. Conclusion KAMINI is a unique 233 U fuelled neutron source facility operating in India. It is providing R&D facilities for neutron
880
S. Usha et al. / Nuclear Engineering and Design 236 (2006) 872–880
radiography, activation analysis and radiation physics experiments. Seven years of operating experience with this facility has been satisfactory. Efforts are underway to enhance utilization of KAMINI for radiation physics and shielding experiments by augmentation of facilities for irradiation. References Dietrich, J.P., 1958. Experimental determination of self regulation and safety of operating water moderated reactor. Geneva Conference Paper. Mohapatra, D.K., 2004. Experimental and theoretical investigations in a thermal neutron research reactor – KAMINI. Doctoral Thesis. Submitted to the University of Madras, India. Mohapatra, D.K., Mohanakrishnan, P., 2000. Moderator temperature effect on reactivity in light water moderated experimental reactors. Ann. Nucl. Energy 27, 969–983. Mohapatra, D.K., Mohanakrishnan, P., 2002. Measurement and prediction of neutron spectra in the Kalpakkam Mini Reactor (KAMINI). Appl. Radiat. Isotopes 57, 25. Mohapatra, D.K., Radha, E., Mohanakrishnan, P., 2004. Theoretical and experimental investigations of reactor parameters in a
U-233 fuelled research reactor. Ann. Nucl. Energy 31, 197– 212. Pasupathy, C.S., et al., 1990. Physics and instrumentation of 233 U fuelled neutron source reactor KAMINI, BARC-1532, p. 160. Radkowsky, A., 1986. Seed and Blanket Reactors, CRC Handbook on Nuclear Reactor Calculations, III. CRC Press, Florida, USA. Ramanarayanan, R.R., Anandkumar, V., Mohanakrishnan, P., Pillai, C.P., Kumar, P.V., Kapoor, R.P., 1999. KAMINI reactor commissioning and operating experience, research facilities and their utilization. In: Proceedings of the IAEA Symposium on Research Reactor Utilization, Safety and Management, Portugal. Ramanarayanan, R.R., Venkataraman, B., Raghu, N., Johny, T., Ramalingam, P.V., Kapoor, R.P., 2002. Utilization of KAMINI reactor for neutron radiography and activation analysis. In: Proceedings of the National Seminar on NDE (NDE 2002), Chennai, December 5–7, 2002. Singh, K., et al., 1980. Adequacy of a time constant concept for the estimation of fuel assembly temperatures in nuclear reactor power transients. In: Paper Presented at the 5th National Heat and Mass Transfer Conference, Hyderabad, February. Technical Specifications for KAMINI operation, 1998. Atomic Energy Regulatory Board, India.