PBMR design for the future

Transcription

1 Nuclear Engineering and Design 222 (2003) PBMR design for the future A. Koster, H.D. Matzner, D.R. Nicholsi PBMR Pty (Ltd), P.O. Box 9396, Centurion 0046, South Africa Received 2 May 2002; received in revised form 1 October 2002; accepted 20 December 2002 Abstract The Pebble Bed Modular Reactor (PBMR) being developed by Eskom and partners in South Africa, has been under development since In the process, it was found necessary to increase the unit power in order to stay within the commercial targets set for the project. This was done in successive steps whereby the basic inherent safety characteristics were always kept uppermost in mind. The conceptual design has moved from a copy of the HTR-Modul developed in Germany to a direct gas cycle design utilising the Brayton cycle. In the process of upgrading the output power, the design first adopted an annular core with a dynamic centre column, which was recently replaced by a solid centre reflector. This design change allows a thermal reactor power of 400 MW and keeps the operating fuel temperature below 1130 C for low fission product contamination of the turbines. At the same time, the geometry of the core structures was kept such that passive cooling to the environment keeps maximum fuel temperatures of the TRISO-coated particle fuel within the safe limits as proven with predecessor reactors like the AVR Elsevier Science B.V. All rights reserved. 1. Introduction The Pebble Bed Modular Reactor (PBMR) under development by Eskom and partners in South Africa has been evolving steadily from a design based on the HTR-Modul of Siemens to a design with a much higher power output and utilising a closed cycle gas turbine for electricity generation. The design utilises the TRISO fuel used in all present HTR gas cooled reactor designs and described repeatedly, like in Lohnert et al. (1988). It also utilises the same concepts of passive safety as used in some previous designs (Weisbrodt, 2003, Doe-HTGR ). Some of the reasons behind the decisions made during the conceptual design phase are described and the plans for the Corresponding author. Tel.: / (cell); fax: address: albert.koster@pbmr.co.za (A. Koster). generation of PBMR power plants that will follow the building and operation of the demonstration reactor are outlined. 2. Safety and economics Safety and economics are uneasy bedfellows and in an environment where even the building of an ultra safe nuclear power plant is regarded with suspicion in many parts of the world, such a concept cannot expect to get the same tax breaks and incentives that other environmentally friendly electricity generating concepts demand to be viable. Thus, the challenge for any new nuclear plant that aspires to be accepted as being free from the risk of a major health threatening accident is to be both that and be competitive with present day generating options like coal and gas fired power stations /03/$ see front matter 2003 Elsevier Science B.V. All rights reserved. doi: /s (03)

2 232 A. Koster et al. / Nuclear Engineering and Design 222 (2003) From the outset the PBMR team set certain safety targets that can be summarised as follows: There shall be no design base or credible beyond design base event either from within the reactor or from external sources that would occasion the need for anyone living near the site boundary to take shelter or be evacuated; There shall be no need for moving mechanical components to ensure that the targets set above are achieved; Exposure of plant personnel shall be significantly lower than the best international values presently being achieved. At the same time the economic goals were formulated as follows: Electricity produced by the PBMR plants shall be competitive with present major electricity generating systems when calculated for the lifetime of all plants and assuming conservative discount rates as well as escalation costs for the various fuels; There shall be no need for a large exclusion zone to enable plants to be built close to populated areas; A short construction time shall allow flexibility in electricity supply planning Attainment of safety goals The danger to the public for any nuclear reactor lies in the fission products contained within the fuel and its cladding. If there is to be no event that may need evacuation of residents near the reactor, then the vast majority of the fission products must remain within the fuel for all credible events down to a very low expected frequency of occurrence. For HTR fuel, i.e. the TRISO-coated particles, this can be virtually guaranteed as long as the maximum fuel temperatures remain below about 1600 C. In order to achieve this there are three well known principles to adhere to: 1. Control the heat production in the fuel by limiting decay heat levels and excessive reactivity excursions. 2. Ensure sufficient heat removal capability (including the options of doing this by passive means, so that the fuel will not exceed the specified limit). 3. Limit the possibilities of chemical degradation of the fuel due to large scale air or water ingress (in this case, the problem is that the SiC coating can be destroyed by corrosion when temperatures exceed 1200 C). These were the basic principles that also applied to the design of the HTR-Modul, and with the choice of a homogeneous core it limited the core size to 3 m in diameter due to the reduced control capability of reflector-based control rods as well as heat removal problems for larger cores. Any design that aims to significantly increase power output above that possible for such a core, and also have the means to control reactivity, must move towards an annular core design. This move was made by the PBMR designers in 1998, and with the original aim of building a lifetime plant the central reflector volume had to be provided by graphite spheres that circulated together with the fuel spheres. These spheres were replaceable if the fast neutron fluence became higher than the expected lifetime. This design made it possible to increase the safe power levels from 200 to 300 MW thermal, even with the high temperatures used in the direct gas cycle. The stated aim for competitiveness of energy produced by the PBMR with other energy sources gave target figures of a cost of between US$ 1000 and US$ 1500 kw 1 installed capacity, depending on where in the world the reactor is to be erected. Using this cost target and increasing accuracy in predicting the as built cost of the whole system, it became clear in 2001 that cost savings by improving and simplifying the design alone would not bring the PBMR within the required cost range. An increase in capacity was called for. Increasing the power level with the existing 302 MW design would lead to unacceptable high accident as well as normal fuel operating temperatures and the decision was made to investigate designs that will allow higher power levels without sacrificing the safety principles. The design changes envisaged and already partly incorporated in the demo plant design are the main subject of this article. However, to understand these changes, some knowledge of the older design is needed and this is provided in the next section Salient data for the 302 MW design In order to understand the limitations of the 302 MW (thermal) design some design details are necessary. In Figs. 1 and 2 is a schematic of the closed Brayton cycle layout with the three shaft configura-

3 A. Koster et al. / Nuclear Engineering and Design 222 (2003) Fig. 1. Simplified diagram of a direct Brayton cycle nuclear power conversion system. tion chosen as the optimal solution for the PBMR as well as the entropy diagram. For the closed Brayton cycle a recuperator is needed to recover heat while at the same time compressing the gas at low temperatures in order to achieve high efficiencies. The helium coolant is circulated by the two high speed turbine compressors through the loop and the usable energy is converted into electricity by the turbine generator set running at 3000 rpm. The cycle is recuperated to retrieve energy and the inlet and outlet temperatures to and from the core are 500 and 900 C, respectively. The entropy diagram for a recuperated three shaft Brayton cycle for the 302 MW design is depicted in Fig. 3. The physical layout of this design is given in Fig. 4 showing the relative sizes of the systems, the recuperator being placed below the turbine generator. This drawing already includes the upgraded core design. Cycle efficiency is dependent on cooling water temperature, pressure, outlet temperature and losses caused by required cooling flows to components in the high temperature region. Fig. 2. Ideal Brayton cycle temperature vs. entropy diagram.

4 234 A. Koster et al. / Nuclear Engineering and Design 222 (2003) Fig. 3. PBMR temperature entropy diagram for 7 MPa main system pressure. The reactor core is depicted in Fig. 5, showing the general arrangement with graphite spheres in the centre and the fuel annulus inside the outer graphite reflector. Between the graphite zone and the fuel, a mixing zone is formed as a result of the formation of graphite and fuel sphere cones below the loading points with some graphite spheres sliding into the depressions between the fuel cones and fuel spheres sometimes bouncing into the graphite region. This mixing zone plays an important role as the neutron flux in that region is highly thermalised and the fuel spheres produce on average more power than those in the fuel region proper. The maximum fuel temperature is coupled to the maximum coolant temperature in the core and is high because of the large bypass flow through the centre graphite pebbles. Fuel and graphite spheres are continually extracted at the bottom, separated and re-fed at the top through nine fuelling and one graphite line. Aside from the bypass through the centre there are also bypass flows in the side reflector due to the reflector structure consisting of 36 columns of graphite blocks joined with keys. The fuelling method requires the pneumatic transport of fuel and graphite spheres to the top of the reactor where they are braked by a counter gas flow in order to limit the impact velocity. This arrangement gives rise to the graphite/fuel mixing zone shown in Fig. 5. The extent and composition of the expected mixing zone was determined through simulation and a depiction of the resulting core configuration is given in Fig. 6, where the centre cone consists of graphite spheres and the surrounding cones are fuel spheres. The thermal neutron flux peaks in the centre and is also higher in the mixing zone than in the fuel zone. This impacts negatively on the maximum fuel temperatures in a depressurisation event as the fuel in the mixing zone provides more energy than the fuel closer to the outside surface. It also reduces the effectiveness of the control elements that are housed

5 A. Koster et al. / Nuclear Engineering and Design 222 (2003) Fig. 4. Physical arrangement of main pressure vessels including turbines. Fig. 5. Core configuration.

6 236 A. Koster et al. / Nuclear Engineering and Design 222 (2003) Core design The task given to the core designer was to design a core that would allow the maximum power to be delivered to the Brayton cycle given that the normal operating fuel temperature should not exceed 1130 C with a coolant outlet temperature of 900 C. Further, the maximum fuel temperature for the design basis depressurised loss of cooling with heat removal through passive means to the Reactor Cavity Cooling System should not exceed 1500 C in a best estimate calculation. These two requirements, together with the obvious one of being able to control and shut down the reactor, to a large extent dictate the shape of the reactor core. For passive removal of the decay heat after a loss of coolant event, the core needs to be long and slender to provide the required radiation surface on the pressure vessel. This heat is transported by radiation and convection from the RPV to the cavity cooling system which has a capacity to passively absorb this heat for more than 72 h. Due to the proven history of the fuel in the German programme, a burnup of average 80 GWd/tU was specified. There are obvious reasons to keep the height of the core to a minimum to save on other construction costs and the core design that emerged is sketched in Fig. 7. This design includes a carbon block layer as insulation to protect the barrel and RPV from excessive temperatures. 4. Turbine design Fig. 6. (a) Computer simulation of pebble loading for dynamic centre reflector design. (b) Computer simulation of pebble loading for dynamic centre reflector design (top view). in the side reflector close to the core. These two factors, as well as the bypass flows in the centre acted as limiting factors to a possible increase in power level other than increasing the outlet temperature. PBMR is not the first HTR project opting for a direct gas cycle system. In Germany, a helium cycle was operated for a short time in Oberhausen and there were other designs in the pipeline when the whole reactor development ended in Germany. Another example of a direct cycle is the GT-MHR being developed by GA and Russian engineers. Previous projects used a standard horizontal turbine separated from the generator by seals. There was also not the knowledge base on gas fired turbine generators available that there is at present due to the fast growth in natural gas fired power stations. One of the crucial decisions was to suspend the turbine with electromagnetic bearings which dictated a vertical placement of the power turbine generator (PTG). This was made possible by the development of large size electromagnetic bearings

7 A. Koster et al. / Nuclear Engineering and Design 222 (2003) Fig. 7. Core calculational layout for 302 MWt reactor model. not available in the 1980s. Nevertheless the experience base of helium driven generators is almost non-existent and this area is receiving a lot of attention from the PBMR team and the selected supplier. Through this partnership it is considered that the risks inherent in the development of such a system have been significantly reduced, even though a few crucial design changes where made in the course of the process as is detailed later Multi-shaft versus single-shaft choice From the time the decision was made to opt for a gas turbine, the question was whether a single- or multishaft design was the better choice. The single-shaft design has some advantage for the load rejection event, where the compressor can act as a brake to limit the rotational velocity. However, the disadvantage of the large size inherent in a design which has to operate at 50 (or 60) Hz was considered to be of greater importance, particularly as it is at present regarded as likely that the high pressure turbine will need maintenance replacement every 6 years. Working with a multi-shaft design entails a different control strategy whereby a number of bypass valves across the recuperator and the compressors are used to compensate for fast transients and load following operations. In Fig. 8, the gas cycle design for the 400 MWt model as described in the next section is shown Increasing the power level The average and maximum operating temperature of the closed gas cycle is considerably higher than previously constructed commercial power reactors (even though the AVR operated for an extended period at between 900 and 950 C). Because of this, insufficient irradiation data for graphite above 700 C are available and accurate lifetime predictions of the inner surface of the graphite reflector is impossible without such data. For this reason, a decision was taken to design the reactor core for replacement of the inner reflector in mid life. This in turn enabled the step towards a replaceable solid centre reflector, thereby eliminating the inevitable mixing zone region where the maximum fuel temperatures were present. In theory, an annular core reactor can be made as large as is needed for the power level desired. The limitations are that the decay heat must be removable by passive means, which demands a steel pressure vessel. The manufacturability and transportability of the pressure vessel then play a decisive role. The PBMR designers have made a conscious decision to limit the size of the pressure vessel to 6.2 m diameter

8 238 A. Koster et al. / Nuclear Engineering and Design 222 (2003) Fig. 8. PBMR 400 MWt cycle diagram.

9 A. Koster et al. / Nuclear Engineering and Design 222 (2003) so that for most possible reactor sites the transport of the components is still possible, even if difficult. Site welding of components is undesirable and should be limited to those components that just cannot be transported in one piece. As the core barrel is almost as large as the pressure vessel, and should be manufactured completely in the factory, this adds another barrier to going for larger reactor sizes. With the limitations mentioned, core design calculations predict a maximum thermal power level of 400 MW for a design that still complies with the stated safety features. A problem with the previous PBMR design, compared to f.i., the block fuel type reactors, is that the control capabilities were limited to the outer reflector region. That by itself was already one of the major stumbling blocks in going for a larger homogeneous core design. The availability of a solid centre reflector immediately offers the possibility of placing control elements in the centre, where their effectiveness is much higher. PBMR have opted to place the cold shutdown system, that is the small absorber spheres, in the centre column, as this provides the position with the highest reactivity effect and will enable cold shutdown to below 100 C using this system alone. As it is deemed necessary to test all important features of a future higher output design already in the demonstration reactor, the core structures design as well as the circulating system design parameters for the planned reactor have been adjusted to enable later upgrades to higher power levels. With the cold shutdown system moved to the centre, the outer reflector can be made of larger blocks to reduce machining costs and shortening the replacement time for the inner blocks. A cross section of the new core structures layout is given in Fig. 9, showing 24 control rod borings and 9 SAS borings in the centre reflector. Fig. 9. (a) Axial cut of reactor vessel and internals showing centre reflector (400 MW design). (b) Cross section of reactor vessel and internals showing defuelling chutes (400 MW design).

10 240 A. Koster et al. / Nuclear Engineering and Design 222 (2003) Fig. 10. Power levels and generator speed for first 100 s after load rejection.

11 A. Koster et al. / Nuclear Engineering and Design 222 (2003) An important change, also partly imposed or allowed by the use of the solid centre reflector, is that the fuel is now extracted through three separate defuelling tubes, each served by its own defuelling machine. Already as a result of value engineering efforts and maintainability concerns, the decision had been made to transport the fuel back to the core outside the RPV in a special fuel shaft that contains the fuelling pipes and some service pipes. The number of fuelling positions is reduced to three which is considered adequate (now that the dynamic centre of graphite spheres is gone) to enable a fuelling cone that is not too high to cause significant neutron flux deviations. A concern with pebble bed reactors is that fuel flow must be relatively even throughout the reac- tor so that no serious azimuthal burnup deviations are to occur as these can lead to long term power distribution deviations not covered by the safety analysis. With the three defuelling lines, a check on possible disturbances in fuel flow is made more readily detectable and fuel circulation can be suspended or adapted until the situation is rectified. With the development of more sophisticated simulation software, it was found that the original three shaft design was, from an efficiency and economic viewpoint, close to the optimal choice. With the planned upgrade to 400 MW the pressure ratios across the compressors need to be increased with the result that it is not possible anymore to keep the resonant frequencies above the operating rotational speed. In a load Fig. 11. Arrangement for turbine compressors and power turbine generator for 400 MWt design.

12 242 A. Koster et al. / Nuclear Engineering and Design 222 (2003) rejection scenario, the generator bypass valves (GBPs) open in 300 ms and electrical output is expended in a resistor bank in order to prevent the generator from exceeding the specified maximum operating speed of the generator. It needs to be mentioned that even if the bypass valve (which consists of eight separate valves in parallel) fails to open in time, the normal control valve system ensures that generator does not exceed its mechanical design limit, thus, no permanent damage would be expected. Fig. 10 shows the simulation of a load rejection event with the actions of the bypass valves and the generator frequency response. For maximum efficiency, the Brayton cycle needs to operate at a pressure that is optimised for the power level chosen. This is one of the reasons why control by helium inventory is used for those power plants that are expected to have to do regular load following. Increasing the power level to 400 MWt (i.e. 165 MWe for the commercial plant) means adjusting the primary pressure to a level of 9 MPa. This increase has led to a change in the pressure ratio of the turbine compressors. For future multi-module power plants, it is expected that load following will not be a major part of plant life and that operating with opened bypass valves to reduce the output to the turbine will be acceptable from an economic viewpoint. The reactor itself can be operated anyplace between 40 and 100% of full power with the available reactivity margin and can also be operated at lower power levels under regulated conditions. A cross section of the power turbine and the turbine compressors is shown in Fig. 11. Originally the complete power turbine generator was supported on electro-magnetic (EM) bearings. Accurately predicting the exact EM bearing sizes needed to balance the PTG is considered a problem due to the need to accurately predict the loads at all modes of operation. Also, no catcher bearings of the required capacity are presently available. There are also other problems connected to having the generator running in helium containing radioactive graphite dust as well as having high tension penetrations through the pressure boundary, and a decision was made to separate the turbine from the generator with a set of dry seals. This means that these now become part of the pressure boundary and that care must be taken that these do not become a weak point in the barrier that prevents helium loss to the environment. Experience up to date indicates that bearings of the size needed can be made and be reliable to the level required to not cause unplanned outages with a frequency that will impact negatively on the availability targets. A detail of the placement of the dry seals in the PTG is given in Fig Building layout The most recent layout of the building for the demonstration reactor incorporating the major features of a 400 MW design, as well as a proposed layout for a power station consisting of eight modules are shown in Figs Fig. 12. Dry seal arrangement between turbine and generator.

13 A. Koster et al. / Nuclear Engineering and Design 222 (2003) Fig. 13. Cut through building at midpoint showing power plant. Fig. 14. Cross section at zero level of power plant.

14 244 A. Koster et al. / Nuclear Engineering and Design 222 (2003) Fig. 15. Layout of a eight-module power plant showing various levels. 6. Status By the end of 2002, all the major systems will have been designed to a level where basic design work can begin bar the completion of a number of thermo-hydraulic analyses. All aspects of importance in a nuclear plant not directly related to system design have also been included, that is radiological protection, maintenance plans, decommissioning, spent fuel storage and transport, site-specific sea water supply, etc. The shareholders have been supplied with a business case detailing the cost estimates, risks and potential

15 A. Koster et al. / Nuclear Engineering and Design 222 (2003) markets, all provided by an independent and credible consulting firm. The prospects for an ultra safe and economically competitive nuclear power plant that can be expanded at the rate dictated by electricity demand are extremely bright. Should permission to proceed be granted early in 2003, then the expectation from an aggressive project plan is that criticality can be achieved sometime in 2008, dependant on funding. An infrastructure of analysis tools manned by 130 engineers and scientists will roughly double over the coming 2 years so that associated tasks like planning the multi module plant and gearing up for design accreditation in the USA can be handled in the same time span. Although most of the detail analysis is performed in South Africa, PBMR itself is also the coordinator and integrator of much design and development work done by suppliers and potential suppliers so that the number of engineers working on the project is more than double the number working in South Africa. 7. Experimental support Although there is a wealth of experimental data from the German programme available, there are sufficient differences between previous designs and the PBMR concept to require some additional testing. The following experimental programmes have been initiated and are either already completed, partly running or in construction. As part of the core design V&V programme, a series of reactor physics experiments were performed in the ASTRA critical facility at the Kurchatov Institute in Moscow, and excellent agreement between experiment and calculation was established. As the core design matures, further tests will be performed at this facility. As a three shaft closed Brayton cycle has never been build before, a micro model of such a system that mimics the PBMR design and control and operating with nitrogen, was recently commissioned at the University of Potchefstroom in South Africa. The model (which has a mass of 35 tonnes) has so far performed flawlessly and the simulation predictions were extremely close to the experimental values (<1% deviation). A Helium test loop is being constructed to test a variety of components like valves, control rods etc. under actual operating conditions. It will be used in the pre-qualification process to ensure that the design complies with requirements. A scale model pebble flow mock-up has operated to test a variety of defuelling concepts and is presently being upgraded to provide verifiable pebble flow parameters for use in both reactor physics and thermohydraulics analysis. A full scale turbine test facility is planned at the premises of the supplier to verify analysis and obtain operational maps for use in simulator studies and qualify the electromagnetic bearings. 8. Conclusion The original design objective of the PBMR was a single module stand alone power plant able to do fast load following. Several modules could be placed on the same site and use a common service building, but would not share other facilities. This option is true for the planned demonstration module to be erected at Koeberg and is suitable for remote locations or countries with an underdeveloped electricity generation system. This solution is not competitive in countries with a well developed electricity infrastructure or where gas prices are low. A second option being developed in parallel is to have a multi module power plant within a common building and sharing some facilities. If the power plant is intended to do only base load operation, such a plant will only need one helium inventory system for 4 8 modules. In common with previous concepts, this arrangement leads to cost savings which, together with the design changes mentioned above, make the concept competitive with almost all other forms of electricity production. This statement is confirmed by recent in depth cost studies and market analysis. The changes made to the design are still compatible with the safety goals described in the beginning. References Lohnert, G.H., Nabielek, H., Schenk, W., The fuel element of the HTR-Module, a prerequisite of an inherently safe reactor. In: Proceedings of the 10th International Conference on HTGR s, San Diego, Paper II.24. Weisbrodt, I.A., Inherently Safety Design Features of the HTR-Module, IAEA-CN-48/125. HTGR, The Modular High-temperature Gas-cooled Reactor (MHTGR) in the U.S., Doe-HTGR