Induced nuclear reactions: processes – systems – and elements – Epi-thermal reactor structures
Reexamination Certificate
1999-09-08
2001-07-10
Jordan, Charles T. (Department: 3641)
Induced nuclear reactions: processes, systems, and elements
Epi-thermal reactor structures
C376S406000
Reexamination Certificate
active
06259760
ABSTRACT:
BACKGROUND OF THE INVENTION
1. Field of the Invention
This invention relates to an apparatus and a method for producing steam from nuclear power for use in generating electricity. More particularly, it relates to system utilizing a fast or epithermal spectrum reactor core immersed in a pool of light water, together with redundant components such as steam generators and reactor coolant pumps, which can be operated for in excess of at least ten years and preferably more than fifteen years, without refueling or access to the internals, such that the system is proliferation resistant. The invention includes operating a nuclear steam supply system at lower average temperature than conventional reactors and with partial boiling in the core to achieve the reliability and efficiency needed for such maintenance-free long life.
2. Background Information
Nuclear steam supply systems (NSSS) employ a nuclear reactor to generate steam which is typically used to drive a steam turbine for the commercial production of electricity. The two most common types of NSSS, the PWR and BWR, utilize light water as the reactor coolant to extract heat from the reactor core. In the PWR (pressurized water reactor), the reactor coolant is circulated through two to four primary loops which include steam generators external to the pressure vessel housing the reactor core. The steam generators utilize the heat in the reactor coolant to convert feed water to steam which is delivered to the steam turbine in secondary loops. The hydraulic head for circulating the reactor coolant in the primary loops is primarily provided by reactor coolant pumps which are also located external to the pressure vessel.
In the BWR (boiling water reactor), the light water reactor coolant boils and the steam is delivered directly to the steam turbine, also typically in multiple loops. Reactor coolant pumps in return piping deliver the condensate of the reactor coolant back to the reactor core to complete the single cycle.
Both the PWR and the BWR are thermal spectrum reactors in which the light water reactor coolant also serves as a moderator. The neutronics of such a thermal reactor require materials with a low cross-section of neutron capture, such as Zircaloy® cladding to contain the fuel. Such reactors need to be refueled, typically about every 15 to 18 months. This requires access to the core for replacement and reshuffling of the fuel rods.
Another type of reactor, the liquid metal reactor (LMR), utilizes a liquid metal such as sodium as the reactor coolant. In order to maintain the separation between the feed water for generating steam and the radioactive primary sodium, an intermediate heat exchanger is mounted within the reactor vessel to heat a secondary sodium fluid which is then circulated through an external steam generator. Such size LMRs utilize a fast spectrum reactor core which can potentially provide a long core life.
A PWR utilizing a fast spectrum reactor core in a typical multiple external loop NSSS has been proposed. The fuel rods were arranged in a triangular, tight lattice configuration, with a pitch (spacing between rods) to diameter (of rod) ratio of less than about 1.1 to achieve the desired neutronics. This core had a proposed life between refuelings of about five years.
It is known to increase the heat transfer area of fuel rods by increasing the surface area such as with fins or lobes on the outer surface. This allows the fuel temperature to be reduced.
Advanced NSSS have been proposed with a primary focus on improved safety. These include a system in which the steam generators for a PWR are immersed in a circulating pool of light water within the reactor pressure vessel, without the need for primary loop piping, thereby eliminating one of the primary concerns in a PWR, the loss of coolant accident (LOCA). This system also uses a slow spectrum reactor core.
There has been little interest in recent years in new nuclear power plants. To break this stagnation affecting the nuclear industry, new reactor designs have been proposed from various sources. Characteristics addressed have been increased safety, reduced cost, reduced construction time, and ease of maintenance. Since it has been projected that whether and when a new reactor will be built, this will occur overseas, possibly in a developing country, the problem of nuclear proliferation has been raised.
While designs have been proposed which address some of these issues, and particularly the issue of safety, neither the presently available nor the proposed systems have adequately addressed the nuclear proliferation issue. The light water reactors, especially, require frequent replacement or reshuffling of the fuel rods, which means that the operator must access the core, creating the opportunity for proliferation. Even the proposed PWR with a fast spectrum reactor core only has a projected life of about five years. Also, while the proposed light water NSSS with the pool configuration removes the risk of a LOCA, the steam generators of a PWR are typically subject to failures. Placing them inside the pressure vessel increases the need to open the pressure vessel for maintenance.
There is a need, therefore, for an improved NSSS and its method of operation.
In particular, there is a need for such an improved system and method with enhanced safety, lower construction costs, lower operating costs, and which requires less maintenance.
There is an additional need for such a system and method which is proliferation resistant.
Specifically, there is a need for such a system and method which can be operated for ten or fifteen years or more, without replacement or reshuffling of the fuel rods, and which in general does not require access to the internals of the pressure vessel for at least that period of time.
SUMMARY OF THE INVENTION
These needs and others are satisfied by the invention, which is directed to a unitary, transportable, assembled nuclear steam supply system, which includes a reactor core immersed in a pool of reactor coolant in an upright pressure vessel. A plurality of steam generators are immersed in the pool of reactor coolant within the pressure vessel. Each of these steam generators has a secondary circuit extending outside the pressure vessel. The plurality of steam generators exceeds the number of steam generators needed to operate the nuclear core at full power. Means are provided to induce a flow in the pool of reactor coolant through the reactor core and the steam generators, and back to the reactor core. The pool configuration eliminates the piping losses inherent in a loop system, in addition to providing the obvious safety advantage that LOCAs and pipe rupture accidents are no longer possible. As the steam generators are more susceptible to failures, redundant steam generators in excess of the number required with the core at 100% power are provided, so that with the failure of a steam generator one of the spare generators can be activated, therefore eliminating the need to open the pressure vessel. This not only enhances overall system reliability, but contributes to the proliferation resistance of the system. While, as will be seen, the system is designed to promote natural circulation, redundant reactor coolant pumps are provided, so that again the pressure vessel does not have to be opened in the event of the failure of a pump.
The flow path of reactor coolant within the pressure vessel is defined by a core chimney which extends centrally upward within the generally cylindrical pressure vessel from the core, which is located generally centrally in the lower section of the vessel. This core chimney forms a central passage above the reactor core and an annular passage between the chimney and the adjacent inner surface of the pressure vessel. Circulation in the pool of reactor coolant is upward through the reactor core and the central passage and then downward through the annular passage to the underside of the reactor core. The plurality of steam generators are immersed in the pool of reactor coolant in the upper section of the pressure
Carelli Mario D.
Green Lawrence
Paramonov Dmitry V.
Zhan Nelson J.
Jordan Charles T.
Mun Kyongtack K.
Westinghouse Electric Company LLC
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