Department of Nuclear Engineering, University of California, Berkeley
Nuclear Engineering - 161 Term Project Report, Fall 1994
Despite the great promise of solar, geothermal, and fusion power, only nuclear fission energy and coal can be relied upon to supply energy in the quantities needed to make this transition possible during the next ten years. In the United States four types of nuclear power plants are expected to play the major roles in nuclear power generation - pressurized water reactors(PWRs), boiling water reactors(BWRs), high temperature gas-cooled reactors(HTGRs), and liquid metal-cooled fast breeder reactors(LMFBRs). Here we will concentrate the discussion on the high temperature gas-cooled reactors(HTGRs).
HTGR is used as an alternative to light water-cooled and moderated reactors. It uses graphite as a moderator and helium as the coolant. HTGR has been used for electric power generation for a long time. The helium coolant enters the reactor at 636F and exits at 1377F. It is possible to employ conventional superheat/reheat cycles found in high performance fossil-fueled power plants becauseof these high temperatures. The overall plant efficiency for HTGRs is 39 percent which compares quite favorably with the 33 percent efficiency obtained with LWR power plants.
In addition to the plant we discussed above, General Atomic has performed design studies on a HTGR employing gas turbines for electric power generation. This plant incorporates a direct cycle because the reactor coolant is expanded through the gas turbine. The turbine-compression units are placed inside the prestressed concrete reactor vessel(PCRV) which enables complete isolation of theprimary coolant. An added feature of this type of plant is the dry cooling can be used because of the higher temperatures for heat rejection. The design parameters for HTGR are listed in Table3.1.
The PCRV is a vertical prestressed concrete vessel with water-cooled steel liners for the internal cavities and their connecting ducks. It is a leak-tight pressure vessel which contains the reactor core, reformers, steam generators,circulators, and primary coolant helium. It also serves as a biological shield for the reactor. The plugs of composite steel and reinforced concrete are the closures for the heat exchanger cavities. They are "breech locked" and bolted into the cavities.
The PCRV has an outside diameter of 33.6 m and a height of 26.8 m. The core cavity is located in the center and is 11.2 m in diameter and 13.8 m in high.The 12 main peripheral cavities, which alternately contain the six reformers and the six steam generators-helium circulator assemblies, are all 4.0 m in doameter. The steam generator cavities are 19.1 m long, while the reformer cavities are 22.0 m in length. There are also three auxiliary cooling system cavities, which are 2.3 m in diameter and 21 m long. All cavities are closed by either reinforced concrete or steel plugs, depending on the size.
The water-cooled and insulated leak-tight steel liners serve to limit helium leakage to a maxium of 1% of the total inventory per year. They are important component for all cavities, ducts, and penetrations. The cooling system maintains concrete temperatures to acceptable limits. The thermal barrier for the ducts and penetrations in the PCRV consists of fibrous blankets with metal cover plates. The thermal barrier for the core support floor, which sees hot helium streaks leaving the reactor core, includes layers of ceramic block on top of thestandard thermal barrier. Another advantage of this type of pressure vessel is its field fabrication which eliminates transportation of 1000 ton steel pressure vessel.
Helium at 700 psia enters the top of core and flows downward through circular holes in the vertical graphite fuel assemblies. The helium exits from the core and passes through the reheater section of six seperate steam generators. The helium then passes passes through the high pressure steam portion of the steam generators and exits into the suction side of the helium circulators. Those circulators are driven by steam turbines which obtain steam from from the exhaust from the high pressure turbine. helium then go back to to the core inlet. Any one of those helium loops incorporating a steam generator and helium circulator may be isolated from the core by valves.
Irradiation tests on HTGR fuel have demonstrated acceptable performance infuel that has been operated to temperature above 1350C and to fast neutron fluences above 8 * 10^21 nvt (E > 0.18 Mev). But developments in all aspects of core materials indicate that there is possibility for extending the temperature and neutron fluence limits even higher than current levels. Additives to kernels to enhance stability under high temperatures are known. Coating strength and retentiveness to metallic fission products may be increased by layers or alloys of SiC or other refractory materials. Recently, a near-isotropic graphite which is much more stable at high fast neutron fluences and high temperatures has been developed for use in HTGR fuel elements.
Straightforward changes in the fuel and in the fuel management, many developed for the Fort St. Vrain HTGR core, have been incorporated to achive higher process temperature while maintaining material limits. Some of the more attractive alternatives which may be implemented without extensive development programare listed in Table 5.2. Based on a commercial coreoperating with a 500C inlet temperature, the increase in process inlet helium temperature resulting from various design option is shown in column 2. Table 5.1 indicates that a number of design alternatives are available which allow hot helium temperature increase at a cost in the neighborhood of 1000C/(mill/kWh).
By using TRISO particles on both fissile and fertile particles, a more dimensionally stable fuel rod is produced. The helium space between the rod and the graphite moderator surrounding the hole increases less rapidly with irradiation. Therefore, the heat transfer resistance is smaller than for the TRISO fissile - BISO fertile rods of the commercial HTGR. Use of TRISO coatings on both fissile and fertile particles has been developed for the FSV core. An additional benefit of all-TRISO coatings is to increase the core's retentiveness for metallic fission products. These benefits are accompanied by increased manufacturingcosts for the TRISO fertile particles and a neutronic penalty gue to more neutron capture in the silicon.
If we decrease the C/Th ratio in the core by increasing the thorium loading, it'll lead to a higher conversion ratio and keeps the transverse power peaking down. Requirements for a somewhat higher fissile loading to maintain reactivity requirements pushes up fuel cycle costs. It is also possible to reduce power density easily. However, the energy costs associated with plant capital costsare more unfavorably affected than the fuel cycle.
Although the numbers in Table 5.1 indicating increases in hot helium temperature for design alternatives are not strictly additive for core designs wherecombinations of these features are used, but helium temperatures in the range 850C to over 950C are attainable by combining several of these alternatives.
In the longer term, even higher hot helium temperatures are possible usingcurrent materials by employing a novel "axial pushthrough" fuel management scheme. Axial pushthrough fuel management is, as the name implies, a scheme for "pushing" the fuel blocks through the reactor core from the top to the bottom. Fresh fuel is loaded at the top of the core and depleted fuel is removed from the bottom. Since HTGRs operate on a 4-yr cycle and the active core is eight blockshigh, fresh fuel os loaded into the top two layers each year while depleted fuel in the bottom two layers is discharged. Thus, the fuel blocks progress down through the core, two layers at a time, staying a layer in each axial position.All fuel in a given axial layer is of the same age.
This axial pushthrough refueling scheme has many advantages, the principleones being no radial age peaking, an approximately ideal axial power profile, reduced the maximum fast fluence, and reduced core pressure drop. On the other hand, the refueling procedure required for this scheme is more complex than thecurrent one.
In summary, currently available core designs, fuel materials, and fuel management techniques can be combined to provide an HTGR core design which can attain helium temperatures to about 950C attractive for many process heat applications. A new fuel management scheme is available which can greatly increase the range of application of the HTGR to about 1100C. Adding to this the prospect for further hot helium increases through improvement in core materials, it is evident that the HTGR core can readily be adapted to high-temperature process applications.
Besides the control rods, which serve also as shutdown rod, a secondary shutdown device is demanded by the recent safety criteria in Germany. Up to now there is no final decision on the design of this device.
(2). Residual heat removal system. The system is designed to remove residual heat after normal shutdown and after all anticipated incidents. Because of the high redundancy of the primary He-loops these are also used for residual heat removal. Residual heat is transfered by means of the steam generators to seperate auxiliary steam-water loops. Steam flows through a throttle-valve to a water-cooled condenser and the condensate is recycled to the steam generator. One auxiliary loop is fed by two steam generators. The whole residual heat removal system, except the cooling tower for the condenser cooling, is installed inside the containment. The capacity of the system is 200% for the loss of coolant accident with a primary gas pressure of 3.3 bar, that means a redundancy of the primary loops of 8 * 25 % (or 12 * 16 %) and of the auxiliary loops of 4 * 50 %.
If intermediate He-circuits are used particular He-water heat exchangers are provided for residual heat removal instead of the steam generators.
(3). Containment. Licensing practice in FRG indicates, that all power reactors must be protected against airplane crash, although the probability of such events is in most cases less than 10^-6 per reactor year. The following assumptionsl are made: A projectile of 7 m^2 cross section hits the containment perpendicularly with a time dependent impact. To achieve a reasonable protection a concrete wall of a thickness between 2 and 3 meters is needed. If one accepted this, it is only a small further step to get a containment which is able to withstand an internal pressure of 3.3 bar and which is gastight as it is necessary to govern a large primary leakage. Such a building can also easily designed against the pressure wave of an external gas explosion.
(4). Underground Siting. Because of presumable safety advantages the concept of underground siting of the reactor building was studied to promote the nuclear safety.
There are two different types of underground constructions. One is completely underground with the top of the reactor building about 10 m below ground level with an earth cover. Another one is half underground with an earth coverof about 10 m thickness forming a hill.
6.2 Safety Assessment of the Reactor
To assess the safety of the reactor we considered the risk of radioactivity release and the risk of shutdown associated with typical accidents. In comparison to other types of reactor the potential release of fission products into the containment or environment is rather low. We only distinguish between accidents with and with no release of fission products. The second kind of consequence is the down time or reduction of reactor power.
The probability of the accidents can be estimated in the following way: Ifthe consequence of an accident follows necessarily from the cause, then the consequence has the same probability as the initiating cause. This is especially true for the consequence "shutdown or power reduction of the reactor".