WO1999045545A1

WO1999045545A1 - Nuclear powered steam expansion engine and a nuclear powered generator with method of operation

Info

Publication number: WO1999045545A1
Application number: PCT/US1999/005448
Authority: WO
Inventors: Claudio Filippone
Original assignee: Swann Wayne E; Claudio Filippone
Priority date: 1998-03-05
Filing date: 1999-03-05
Publication date: 1999-09-10
Also published as: EP1074024A1; AU3545699A; WO1999045545A9; WO1999045545A8; EP1074024A4

Abstract

A nuclear power generator having a nuclear fuel assembly which is periodically adjacent first to a first nuclear reactor containing fresh nuclear fuel where the fuel assembly achieves a supercritical state, and then adjacent to a second nuclear reactor containing spent nuclear fuel where delayed neutrons resulting from the supercritical state of the fuel assembly cause a supercritical state in the spent fuel thereby releasing heat. A nuclear powered steam expansion engine capable of converting nuclear energy to mechanical work by periodically bringing into proximity two nuclear fuel assemblies within a piston and cylinder assembly, thus making the fuel assemblies successively supercritical and subcritical in cyclical fashion, with associated cyclical heating, periodically injecting water into the resultant heated cylinder, which water is converted to steam which expands, drives the piston, and produces mechanical work, recycling the expanded steam by adding additional heat during an exhaust stroke, and using the recycled steam to drive a turbine system.

Description

NUCLEAR POWERED STEAM EXPANSION ENGINE AND A NUCLEAR POWERED GENERATOR WITH METHOD OF OPERATION

BACKGROUND OF THE INVENTION

The present invention relates to a nuclear power generator capable of using fresh or spent nuclear fuel to generate power. In its various embodiments, the generator provides an apparatus for moving an assembly of fresh nuclear fuel into and out of proximity to a second nuclear fuel assembly so that the moving assembly experiences consecutive subcritical and supercritical states. The result is both a cyclical release of power by the fresh fuel assemblies, and the periodic availability of a source of delayed neutrons, which can be used to elevate an assembly of spent nuclear fuel into a supercritical state for additional power generation.

The ability of delayed neutrons to elevate an assembly of spent nuclear fuel to a supercritical state has important economic consequences. Spent fuel is available in abundance, is difficult to dispose of, and is generally not used to generate economically useful power. This invention offers the prospect of releasing large quantities of economically useful energy from what is otherwise a dangerous waste product.

Among the embodiments disclosed is a highly efficient Rankine-type engine capable of producing thermodynamic work as a result of the fast

expansion of steam inside a piston-cylinder(s) assembly in which the exhaust steam is recycled by expanding it through a conventional turbine.

In recent years there has been an increasing interest in developing alternative methods for converting heat into work by using optimized Rankine engines. However, in the conventional Rankine engine, unavoidable irreversibilities are associated with each of the sub-systems characterizing the thermodynamic cycle. In particular, the most important sub-system dictating the Rankine engine's overall energy-conversion efficiency is the expansion device, i.e. turbine, and the steam admission valve system controlling the flow of the working fluid into the expander. In general, the expansion devices can be classified into two categories: low-speed systems such as positive-displacement devices, and high-speed systems such as turbines. Conventional turbines are usually not suitable for Rankine power cycle application owing to their low efficiencies when operating with the Rankine constraint. Furthermore, when turbines are utilized their relatively high operating velocities imply the use of an intermediate coupling gearbox between the turbine shaft and the commonly used driven electric generators, thereby decreasing the efficiency even further. For rating speeds below 5,000 RpM, a Wankel-type expander, considered a positive-displacement device, shows the highest efficiency at operational speeds which allow the direct coupling of the driven equipment (electric generators, pumps, etc), thereby avoiding the gear-box inherent irreversibility production and a concomitant loss of efficiency. However, there are still considerable problems associated with a Rankine-modified Wankel expander if a current, commercially available Wankel is to be used. In fact, several problems, including lubrication of the rotating piston, internal-leakage losses, cracking of the apex seals, and the unavoidable fluid friction losses in the intake valve with consequent wear of the valve material itself, have to be solved before a reliable Wankel-type expander can be commercially used. An objective of this invention is to provide a special expander connected through a valve system to a conventional turbine. The special expander is characterized by the integration of the boiler into the expansion chamber of a conventional piston-cylinder assembly. In this manner, the fluid friction losses through the control valve system and the efficiency loss associated with the heat addition process in the conventional boiler are reduced. One goal of this invention is to provide a heat-work conversion engine with efficiency greater than that of conventional systems. Assuming steam is the working fluid, a quasi-ideal situation would be obtained if the boiler and expander would become a single component and the actual heat-work conversion would occur directly and simultaneously in the expander, thereby eliminating the conventional boiler, piping, and valve systems for the control of the steam flow.

In the nuclear powered steam expansion engine (NPSEE) the piston- cylinder system is alternately vented to a conventional turbine, thereby recycling that fraction of expanded steam that would be wasted. The NPSEE can be thought of as a highly efficient stationary power station which is capable of reducing the rate of environmental degradation per unit of energy produced.

SUMMARY OF THE INVENTION An object of the present invention is to provide an apparatus, and a method, of ultimately producing electric power with a high efficiency, thereby overcoming the low efficiency characteristics of conventional power plants. It is a further object of the invention to provide an apparatus for generating nuclear power using spent nuclear fuel by transferring an assembly made of fresh fuel between a nuclear reactor containing fresh fuel, where it achieves supercriticality and a reactor containing spent fuel which is thereby raised to a supercritical state. A variety of apparatus are presented which use the time-variant transition of the moving fuel assembly from subcriticality to supercriticality in the generation of power. These include externally driven hydraulic and mechanical devices that impart periodic motion to the moving fuel assembly. Also included is a nuclear powered steam expansion engine that moves the fuel assembly using internally generated mechanical energy. This mechanical energy is converted internally from the heat of the nuclear reaction itself.

The nuclear powered steam expansion engine (NPSEE) includes high pressure liquid water forced inside specially designed heat cavities hydraulically connected and sealed to a piston cylinder assembly. Considering only one cavity, the high pressure liquid water enters the cavity through a high-pressure injector(s) that opens in synchronism which the position of the piston and as a function of the relative stroke. During the water injection, or the injection of any fluid with certain thermodynamic and neutron moderation characteristics, the piston is found a few crankshaft degrees ("crank") below its top dead center (TDC) position, while the exhaust valve is kept closed. As the water is injected, heat is transferred from the surfaces of the heat cavities to the fluid, changing it thermodynamic state. Now the piston is found in its TDC position and the pressure inside the heat cavity increases rapidly. As a consequence of the increased pressure, the piston is forced toward its bottom dead center (BDC) position, thereby producing the power stroke. When the piston reaches the BDC position, the exhaust valve opens and vents steam to a conventional turbine system through a steam expansion valve. During the exhaust stroke, the piston moves upward, forcing steam through the heat cavities in which heat is transferred to the steam one more time. In this manner, superheated steam is available for further expansion in the turbine(s) producing additional work. Once in the condenser, heat transfers from working fluid at a constant pressure in the same manner as for conventional Rankine cycles. The power cycle starts again when the exhaust valve closes and the piston is re-positioned a few degrees below the TDC position re-activating, through electromechanical links, the high-pressure injectors. Overall, the cycle is completed in two strokes, the power stroke and the exhaust stroke corresponding to exactly one 360 degree rotation of the crankshaft.

In such an engine the typical Rankine cycle processes of heat addition in the boiler, and the fluid expansion in the expansion device, occur simultaneously. In this manner, the rate of entropy production associated with the two-step processes is significantly reduced, thereby increasing the efficiency. Furthermore, the exhaust product of such an engine is superheated steam that is re-utilized, i.e. recycled, by letting it expand in a series of conventional expanders.

From a neutronic viewpoint, the temperature on the surfaces of the heat cavities changes as a function of variable fission rate, which is mainly affected by the motion of the piston. The piston can be considered as a source of neutrons (mainly delayed) which are emitted even when it is found in its BDC position. Therefore, by utilizing this characteristic for graphite moderated reactor, located under the water moderated reactor, it is possible to recycle spent fuel since it still contains a considerable fraction of natural uranium.

The above and other objects, features and advantages of this invention will be clear from the following description of the preferred embodiments when the same is analyzed in conjunction with the accompanying drawings.

BRIEF DESCRIPTION OF THE DRAWINGS Fig. 1 is a sectional view of an embodiment of the nuclear powered steam expansion engine in accordance with the present invention. Fig. 2 is a sectional view of a simplified representation of the high pressure fluid injector(s), taken along line l-l and V-V of Fig. 1. Fig. 3 is a sectional view of a simplified representation of the cavity containing nuclear fuel assemblies, taken along lines Ill-Ill and IV-IV of Fig. 1. Fig. 4 is a sectional view of the transfer cavities placed in the piston with their relative inlet ports exposed to the fluid injectors.

Fig. 5 is a simplified sectional view of the core of the NPSEE in which sets of control rods are utilized in the same manner as for conventional pressurized water reactors (PWRs).

Fig. 6A is a fragmentary perspective view of the piston, piston-supported fuel assembly, and cylinder with the piston near top dead center. Fig. 6B is a vertical section through the center of the piston and piston- supported fuel assembly of Fig. 6.

Fig. 6C is a fragmentary section of the injectors, cylinder, and piston-supported fuel assembly taken along line 6C-6C of Fig. 6B.

Fig. 7 is the model of fissile material rich piston that upon insertion inside a fissile material rich cylinder that overcomes the critical condition to produce heat, steam and ultimately work.

Fig. 8 is the analogue of the NPSEE configured for the recycling of fuel to be positioned in the graphite-moderated reactor located under the water moderated reactor.

Fig. 9 is a schematic of the thermodynamic processes occurring in the NPSEE to form the highly efficient NPSEE thermodynamic cycle. Fig. 10 is a T-S diagram of the NPSEE thermodynamic cycle in which two reference crankshaft speeds are considered.

Fig. 11 is a fragmentary perspective view of the nuclear power generator having both fresh fuel and spent fuel reactors according to the invention. Fig. 12 is schematic view showing a hydraulically actuated apparatus for moving fuel assemblies between a reactor containing fresh nuclear fuel and a reactor containing spent or natural nuclear fuel. Fig. 13 is a schematic view showing a mechanically actuated apparatus for moving fuel assemblies between a reactor containing fresh nuclear fuel and a reactor containing spent or natural nuclear fuel.

DESCRIPTION OF THE PREFERRED EMBODIMENT A first embodiment of the nuclear powered steam expansion engine proposed is described with reference to Fig. 1. As shown in this Figure, the steam expansion engine apparatus has a plurality of units: 1 , 3, 14, 7, 18 ,20 and 22, which are connected in series through connecting pipes as to form the closed loop of a thermodynamic power cycle. The steam expansion apparatus of the present invention has a high-pressure pump in which de-mineralized and de-ionized water is pressurized into a discharge tube to a pressure level such that a series of spring loaded valves (injectors) open, spraying water droplets into high temperature cavities in which water flashes to steam. The rapid expansion of steam causes a pressure increase inside the cylinder-piston assembly. Water is the working fluid considered in this design but any other fluid having the proper thermodynamic, physical and neutron moderation characteristics can be utilized.

With reference to Fig. 1 , unit 1 represents a high-pressure pump that is characterized by one inlet 29 and one exit 2. At the inlet 29, the fluid is tin the liquid state at the post-reactor 14 conditions. At the exit 2, the fluid is at the high-pressure liquid state with a temperature not much different than the inlet temperature. Furthermore, the thermodynamic state of the fluid at the inlet 29 of pump 1 is characterized by a pressure of about 80 bars and a temperature of approximately 230 C. The saturation temperature at this pressure is about 295 C, therefore, the fluid state is found in the sub-cooled liquid region. In general, the pressure of reactor 14 is kept at the constant value of 80 bars by utilizing the pressurizer 38 that operates in the same manner as a conventional PWR pressurizer. Pump 1 has the purpose of changing the pressure of the liquid water at the exit 29 of the reactor 14 from the reactor pressure, about 80 bars, to a pressure of about 90 bars (1323 psi), which is the minimum, pressure value necessary to open the injectors 3,2-1 a and 24b. Compressor 1 is mechanically linked, through the gear system 10, to the crankshaft 26 so that the rotation of the compressor impeller blades 27 is synchronized with the position of the crankshaft 26. In this manner the fluid is compressed in a pulse-like fashion, as in the case for Diesel injectors pumps.

At the exit of compressor 1 , the fluid is cyclically pressurized into the discharging tube 2, which is hydraulically connected to the discharging tubes 25 and 25b and ultimately connected to the body of a series of high pressure injectors 3, 24a and 24b. The pressure at the discharging tube 2 is equalized for all of the high pressure injectors through the hydraulic connection 25a and 25b but, if necessary, by changing the injectors' spring constant, it is possible to delay the opening of one series of injectors with respect to another series of injectors so as to control the rate of expansion of the fluid in different locations of the expansion chamber 36.

As shown in Fig. 2, 3, and 4, the water injectors 3, 24a and 24b are spring-loaded valves (these representations are not to scale). As shown in Fig. 2, inside the body 3 of the high pressure injector there is a calibrated spring 4 mechanically connected to an injection nozzle valve 5. When the pressure inside the injector bodies 3, 24a and 24b overcomes a preset value, which is a function of the position of the camshaft 28, the injection nozzle of valve 5 opens. The injection nozzle valve 5 is contained in each series of injectors 3, 24a and 24b. As soon as valve 5 opens, the pressure in the discharging tube 2, 25a and 25b drops rapidly, causing valve 5 to close again. But since the compression phase is still active, and will remain active for a duration of time determined by the camshaft 28, the pressure in the discharging tubes rises, causing valve 5 to open again. This process results in an oscillation depending on the injector dimensions, spring constant, and operating pressure of the injection nozzle valve 5. The oscillation persists during the whole injection phase determined by the compressor 1 and the position of the impeller blades 27 connected to the gear system 10 and the camshaft 28. The injection duration refers to the degrees of crankshaft rotation, and thereby is a function of the RpM of the crankshaft.

With reference to Fig, 1 , 2, 3 and 4, a certain amount of water, depending on the calibration of the spring constant 4 and the angle at which the crankshaft 28 is turned for the injection phase is sprayed into the gap 6, into the internal cavities 23, and into the internal cavities 34 of the nuclear fuel assembly 13. Once the fluid is injected, its thermodynamic state changes very rapidly, going from a state of pressurized sub-cooled liquid to a two-phase liquid-vapor state and finally to a superheated state. This thermodynamic change induces a significant increase of the pressure inside the piston 12 and cylinder assembly 33.

A general discussion will now be given of the basic principle upon which the temperature of the heat cavities increases as a function of the speed and position of piston 12 carrying fission material in the nuclear fuel assemblies 13. Since the piston system 12 carries a certain amount of fissile material contained in the fuel assemblies 13, it inserts a certain value of nuclear reactor criticality every time it enters the reactor 14. Nuclear reactor criticality indicated with the symbol "k", is the conventional term utilized to define the ratio of neutrons generated at a given time to the number of neutrons generated at a pervious time. In other words, it provides a measure that defines the rate of fission reactions. As shown in Fig. 5, the reactor is in the super-critical configuration when, in its whole, the nuclear fuel assemblies 13, 14 and 7 form a generally contiguous mass. In particular, this situation is represented in Fig. 7. In fact, by combining the amount of nuclear fuel masses contained in the nuclear fuel assemblies 13, 14 and 7, as shown in Fig. 5 and qualitatively represented in Fig. 7, the total amount of mass overcomes the value for which the reactor becomes critical. To control and adjust the mean value of the criticality oscillation caused by the motion of the piston start up and during normal operation, a certain number conventional control rods 31 is required.

During normal operation, the reactor formed by the nuclear fuel assemblies 13, 14 and 7 is cyclically found in a slightly super-critical condition subsequent to a sub-critical condition depending on the position of piston 12 and the control rods 31. The super-critical condition (k> 1) corresponds to the TDC position of piston 12, while the sub-critical condition is at its maximum, in absolute value, when piston 12 is found in the BDC position. The control rods 31 represented in Fig. 5 are utilized in the same manner as conventional control rods are utilized in PWRs. Therefore, as represented in Fig. 5, at the reactor start-up the control rods 31 will be found inserted into the guiding tubes 16 and withdrawn gradually during normal operation and/or depending on fuel burn-up.

Depending on the overall nuclear positioning, material, and reactor geometry, when the piston 12 is found in the BDC position, the negative reactivity insertion can be designed such that it is much higher, in absolute value, than the positive reactivity insertion that occurs when the piston 12 is found inside the reactor 14 and corresponding to the piston TDC position.

Overall, the surfaces of the heat cavities of fuel assembly 7, the reactor fuel assemblies 14, and the nuclear fuel assemblies 13 positioned on the piston 12, experience a pulsed-like temperature increase as a result of nuclear fission. With reference to Fig. 4 and Fig 6, and assuming also that the initial position of the piston 12 is found a few crankshaft degrees below the TDC position and is moving upward, the fluid inlet ports 32 become gradually aligned with the fluid injectors 24b.

The fluid injection phase starts when the inlet ports 32 in Fig. 1 are exposed to the outlet of the fluid injectors 33. At the same time, the camshaft 28, synchronized with the position of the piston 12 through the mechanical link 11 , sets the rotation of blades 27 of compressor 1 for the injection phase. During this whole phase the exhaust valve 8 is forced closed due to the action of the spring 130, as shown on Fig. 4, which is relatively uncompressed due to the position of the mechanical link 9.

The piston 12 is still moving upward and before it has reached the TDC position, all of the injectors 3, 24a and 24b have completed their fluid injection phase through the discharging tubes 2, 25a and 25b. The initiation and duration of the injection phase is a function of the position of crankshaft 28. When the piston 12 is found in the TDC position, it inserts a positive reactivity resulting in a very rapid increase in the temperature of the heat cavity surfaces. The condition is timed and synchronized with the initiation and termination of the fluid injection phase.

Due to the steep gradient of temperature found in the heat cavities 23 (Fig. 3) and 34 (Fig 4) and in the gap 6, the fluid droplets injected experience a rapid heat transfer from the surfaces of the heat cavities. The result is a rapid expansion of the fluid with a subsequent increase of the system piston-cylinder pressure. Since the pressure increases as the injection phase starts it reaches its peak a few crankshaft degrees after the termination of the injection phase. Then the piston 12 and the fuel assemblies 13 are pushed away from the reactor 14, forcing the piston toward the BDC position and imposing the condition for which the value of k is less than unity, thereby making the overall reactor sub-critical again. This situation is the equivalent of a rapid insertion of negative reactivity. In this case, the prompt neutrons rapidly die out, leaving the reactor under the effect of the delayed neutrons generated by the nuclear fission reactions of the precursors.

When the piston 12 is pushed outside of the reactor 14, it produces the power stroke in which work can be utilized at the crankshaft 26. As shown in Fig. 4, once the piston 2 reaches the BDC position, the exhaust valve 8, through the mechanical link 9, (synchronized with the crankshaft 26 with a ratio 2:1), is now forced to open venting relatively high pressure fluid vapors to a steam expansion valve 17 (Fig. 1).

Because of the momentum gained in the previous power stroke the piston 12 is now moving upward forcing the steam through the heat cavities 23 (Fig. 3) again. In this manner, the steam is recycled and experiences another heat addition process, increasing its energy content. Overall, the steam is forced through a steam expansion valve 17 after which it expands inside a conventional set of turbines 18 producing additional work at the turbine shaft 19 (Fig. 1.) As the piston 12 enters the reactor 14, it adds positive nuclear reactivity, leading the reactor to the super-critical condition and re-creating the conditions for a new cycle.

Finally, the NPSEE cycle starts again as soon as the piston 12 is found a few crankshaft degrees before the TDC position. At this point the exhaust valve 8 closes and the fluid injection phase starts again. From the neutronic point of view, when the piston 12 approaches the TDC position it inserts a new wave of prompt neutrons superimposed on a reactivity condition dominated by the delayed neutrons generated in the previous stroke. In general, if nuclear feedback such as moderator coefficient, Doppler effect and others are not present, the reactivity would increase in a sinusoidal fashion superimposed on an exponential reactivity insertion. However because of the combination of natural and designed feedback, the exponential power rising effect can be stabilized to a steady-state average power. Then the reactivity will be oscillating as a sinusoidal function of time (directly proportional to the non-harmonic motion of the piston) superimposed on an average reactivity value. In particular, the amplitude of the sinusoidal-like reactivity insertion is determined by the position of the control rods 31 and the speed at which the piston cyclically inserts the fissile material 13 inside the reactor 14 and 7. In fact, the wave form of the time dependant reactivity is determined by the non-harmonic motion of the piston 12 and its approach to the internal part of the reactor 145.

As shown in Fig. 1 , and Fig. 5, another compressor, unit 22, is included in the design. Here, compressor 22 represents a pump for the cooling of the nuclear fuel assemblies placed in the reactor 14 and 39, in Fig. 3. The coolant in this section of the reactor is the same fluid utilized for the expansion process, but, to improve safety, it is possible to separate the two parts of the overall reactor constituted by sections 7, 13 and 14 into two independent hydraulic circuits. Overall, the methodology and the designed safety features of the steam expansion engine can be the same as those applied for conventional nuclear reactors. Whit reference to Fig. 1 , it is assumed that the reactor 14 is built with the same concepts utilized in the design of conventional PWR/BWR reactors.

In general, and again with reference to Fig. 1 , it is possible to distinguish two vessel structures, the primary vessel 35 and the secondary vessel 37 surrounding the primary vessel. Furthermore, the primary vessel structure surrounds the head of the fluid expansion chamber 36, which is part of the cylinder 35, and surrounds the fuel assemblies 7 so as to form the gap 6 in which the fluid is forced to an intimate contact with the hot surfaces of the fuel assemblies. The internal structure of both primary and secondary vessels consists of fuel assemblies, instrumentation, thermal shielding and so on a suited for conventional nuclear reactors. In the NPSEE, the reactivity can be controlled by moving the control rods 31 (Fig. 5), by changing the concentration of neutron absorbers diluted in the working fluid which is the moderator itself, and by changing the speed at which the piston 12 approaches and enters the reactor 14.

An additional inherent safety feature offered by the design is related to the time dependent geometry of the steam expansion engine. Since the reactor is formed by moving parts, it implies a reactor sub-critical condition whenever the piston 12 is found in the BDC position. In this case, the subsequent negative reactivity inserted in the neutronic system can be such that the overall reactor shuts down automatically. Therefore, by locking piston 12 at it bottom dead center position, the reactor remains in a shut-sown condition even if the rods 31 are completely withdrawn. In other words, the design guarantees the shut-down condition even in the absence of control rods 31 and the chemical neutron absorbers diluted in the working fluid. The overall reactor can be seen as composed by a fast neutron region, located in the areas adjacent to the heat cavities, such as the areas surrounding the gap 6, the fuel assemblies 13 and 7, and a thermal-neutron region characterized by the fuel assemblies located in the reactor 14. As mentioned, the working fluid, utilized for the expansion process, can also constitute the media needed for the neutron moderation, but this is not a limitation. In fact, the primary vessel 35 can be hydraulically independent of the secondary vessel 37, and the fluids utilized in the two hydraulic systems can be different in nature. However, with the purpose of simplifying the design, if the working fluid is also considered to be the moderator owing to its neutron scattering characteristics, it becomes clear that the nuclear fuel contained in the heat cavities 23 (Fig. 3) and 34 (Fig. 6) is exposed to fast neutrons. The neutron scattering in these regions occurs with steam whose low density decreases the probability for neutron-hydrogen interaction.

Furthermore, the neutron mean free path in these regions of the reactor has an order of magnitude comparable to the gap distance 6, therefore neutrons in these areas of the reactor cannot be slowed down to energies in the thermal range. On the other hand, the rest of the reactor, and in particular the nuclear fuel contained in the fuel assemblies 14, can be treated with the methodology applied for conventional thermal reactors.

As shown in Fig. 8, with the purpose of recycling spent fuel 40, the overall NPSEE reactor can be composed of three parts: the pistons, the pressurized vessel surrounding the pistons, and a graphite moderated thermal reactor located above the crankshaft 39. Therefore, the pistons 13 can been seen as a source of neutrons (mostly delayed) which can cyclically inserted insider the graphite moderated reactor 99. The fuel utilized in this section of the reactor is characterized by natural uranium contained in the spent or un- enriched nuclear fuel rods of assemblies 40. In other words, the fuel utilized in the water moderated reactor 14 can be recycled in the graphite moderated reactor 39. This part of the reactor, due to its physical characteristics, would never be able to reach the critical condition for which 1-1 , but by providing an additional and significant source of neutrons the critical and supercritical condition can be achieved again (consider that in typical spent fuel rods a significant percentage of ²³⁸U is still present). In this manner, the NPSEE design provides a valuable feature since it allows the recycling of spent fuel by allowing a normally sub-critical natural uranium reactor to become supercritical through the insertion of neutrons originated when the piston is in the TDC position. Therefore, the delayed neutrons emitted when the piston is in it lower position, approaching and leaving BDC, and leaking out of the piston boundaries would enter the graphite moderated natural uranium reactor 39 (Fig. 8). These neutrons have an enhanced mean free path due to the scattering characteristics of graphite and will be absorbed in a ²³⁸U isotope or a non fissile isotope with large absorption cross sections in the energy range of delayed neutrons (0.2- 06MeV). In particular, to decrease the neutron leakage probability, the radius of the graphite moderated reactor 39 is in the range of 3 to 4 meters and the number of pistons utilized can be varied.

To summarize, in Fig. 8 the NPSEE operates as a result of the contribution of two reactors having different characteristics. The reactor represented by the fuel assemblies 14 is a pressurized water moderated nuclear reactor that changes in criticality as a result of motion of the pistons 13. On the other hand, the reactor represented by the fuel assemblies 40 located in the graphite-moderated reactor 39 represents a feasible way to further utilize spent fuel. In this case, each piston 13 is cyclically inserted inside the water- moderated reactor 14 (Fig. 8) while it exits the graphite-moderated reactor 39. During the following stroke, each piston exits the water-moderated reactor 14 and enters the graphite-moderated reactor 39. This situation is repeated every 180 crank⁰. Therefore, the fission reactions occurring in the fuel assemblies contained in the pistons 13 produce fission fragments which will decay by emitting delayed neutrons. Since the pistons 13 spend 50% of the time outside the water-moderated reactor 14 (or inside the graphite moderated reactor 39) a significant quantity of neutrons is emitted when the piston is moving toward the BDC position. Normally, depending on geometry, the critical condition for large natural uranium reactors (non-enriched or spent fuel) is fairly close to unity, making such reactors impractical. However, if an additional and significant source of neutrons is cyclically introduced in the neutron balance of such reactors, the overall criticality can be greater than unity, thereby leading these reactors to the super-critical condition again.

In Fig. 9 and fig. 10, the NPSEE thermodynamic cycle is represented. Processed 1-2,2-3 and 3-4 correspond to the compression/expansion processes occurring in the piston/cylinder(s) expander when valve(s) 8, in fig. 1 and/or Fig. 8, is kept close. Considering only one piston as is shown in Fig. 1 , as the power stroke terminates the mechanical links presented by systems 9 and 10 open the exhaust valve 8 venting the exhaust steam to the turbine 18.

This process 4-5 in Fig. 9 an Fig. 10 characterized further heat addition during the exhaust stroke. Processes 5-6, 6 7 and 7-8 in Fig. 9 represent the steam expansion in the turbin 18, the heat recovery in the heat exchanger 21 and the condensation process in the condenser 20 as is shown Fig. 1 and 5. The compression process 8-0 in Fig. 9 and Fig. 10 is achieved by the high pressure pump 22 represented in Fig. 1 and Fig. 5. In Fig. 9, process 0-01 is a heat addition process occurring at the heat exchanger 21 shown in Fig. 1 and Fig. 5.

In Fig. 9, process 01-02 represents the heat addition process occurring in the reactor 14 prior to the pressurization in pump 1 as is shown in fig. 1. In the case the NPSEE is configured for spent fuel recycling as is shown in Fig. 8, process

01-02 represents the heat addition of the pressurized water (pressurized through pump 22) which cools down the graphite moderated reactor 39 and the water moderated reactor 14 before exiting at the exit 29 as is sown in Fig. 1 and

Fig. 8. Process 03-1 in Fig. 9 and Fig. 10 represents the heat addition processes occurring in the heat cavities 23 (Fig. 3), and the heat cavities 7 and

34 shown in Fig. 4, Fig. 5, Fig. 6 and Fig. 8. In general, the thermodynamic cycle of the NPSEE is a sequence of twelve processes in which the expansion of steam occurs alternately in two expansion systems, the piston-cylinder assembly 13-35, and the turbine(s) 18. The efficiency of this cycle is strongly dependent on the amount of water injected through the injectors 3, 24b and

24a. If the maximum temperature allowed on the surfaces of the heat cavities 7 and 34 is 720 C the maximum NPSEE cycle efficiency is about 56% with a

Carnot efficiency of about 68%. Since the amount of water injected depends on the number of crank set for the injection phase and the number of RpM, the efficiency mainly varies as a function of the maximum temperature allowed on the surface of the heat cavity assembly 7 and 34. the crankshaft number of RpM and the duration of the injection phase set arbitrarily by adjusting the position of camshaft 28 which determines the motion of the impeller blades 27 of compressor 1 (Fig. 1).

In particular, to emphasize the effect of the changing number of RpM on the overall cycle, as is shown in Fig. 10, for RpM=100, state (2) and sate(3) are still in the two-phase liquid vapor region. As the number of RpM is increased, less water is injected and states (1'), (2') and (3') shift to the right toward increasing entropy regions. This effect is shown in detail in Fig. 10 where state (1) is relative to RpM=100, while state (1') is relative to the optimized number of RpM which yielded to the highest efficiency (in this case RpM=227).

To summarize, as the angular velocity of the NPSEE crankshaft is increased, and by maintaining the number of crank for the injection phase constant, less water is injected through the water injectors 3, 24a and 24b. This leads to an increased area under the NPSEE T-S cycle (Fig. 10), thereby yielding higher efficiencies under the imposed temperature constraint (the temperature on the surface of the fuel cladding material cannot be greater than 720 C for safety reasons).

Fig. 11 illustrates a more general embodiment of the reactor, described above, which can generate useful energy with spent or un-enriched nuclear fuel. The nuclear power generator shown includes a first nuclear reactor 70 that is referred to as the giver reactor. This giver reactor 70 has giver nuclear fuel assemblies 72 that include fresh or enriched nuclear fuel. The generator also includes a second nuclear reactor 74 that also has fuel assemblies 76. The second nuclear reactor 74 is referred to as the taker reactor, and the taker fuel assemblies it contains include spent or un-enriched nuclear fuel. The taker nuclear fuel assembly 76 is disposed adjacent to the giver fuel assembly 72.

It should be noted that the giver reactor may be a water moderated reactor, a heavy-water moderated reactor, a graphite moderated reactor, or a reactor moderated by a combination of the above. The taker reactor is generally a graphite moderated reactor, although graphite equivalent materials or combinations of materials may be used.

A plurality of third nuclear fuel assemblies 78, which also include fresh fuel and are thus net givers of neutrons, are supported for movement between a first position 80 near the giver fuel assembly and a second position 82 near the taker fuel assembly. These third nuclear fuel assemblies 78 are first disposed adjacent to the first nuclear fuel assembly 72 at position 80 and spaced from said second nuclear fuel assembly 76 at position 82. In this configuration, the first nuclear fuel assembly 72, and the third nuclear fuel assembly 78 are supercritical. At the same time, the second fuel assembly 76 is subcritical. Some time later, the third nuclear fuel assembles 78 are disposed adjacent to the second nuclear fuel assembly 76 at position 82 and spaced from the first nuclear fuel assembly 72. In this configuration, the first nuclear fuel assembly 72 is subcritical and the second nuclear fuel assembly 76 is, at least for a time, supercritical.

The motion of the assemblies 78 may be synchronized to insure that the position of each is out of time and position phase with one or more other assemblies, so as to minimize time variation in the power or heat released by the reactors 70,74.

As specifically shown in Fig. 11 , since the movement of each rod 98 is independent it is possible to cycle their insertion inside reactor 70 or 74 as desired so as to generate a smooth transition from sub-critical to super-critical throughout the two reactors. The mechanism that transfers assemblies 78 from one reactor to another is independent and does not require steam as a driving force as it was necessary in the reactor configuration of Fig. 1 and Fig. 9.

As illustrated in Figs. 11 , 12, and 13, the means for moving the third nuclear fuel assemblies 78 between the first position 80 near the giver reactor 70 and the second position 82 near the taker reactor 72, need not be a steam engine. Any appropriate means, such as mechanical, hydraulic, pneumatic, electromotive, etc. would do. An exemplary embodiment utilizing hydraulic actuation is illustrated in Fig. 12. This embodiment shows an arrangement including a hydraulic power source 90, valve means 92 which may include a controller 94 to periodically operate the valve means, connecting piping 96, a pair of moving fuel assemblies 78, each within a cylinder 100, and a reservoir 102 of hydraulic fluid 104 capable of transferring energy from the hydraulic power source 90 to the moving fuel assemblies 98. In this embodiment, fluid 104 is forced through the piping 96 and valve assembly 92 by the power source 90 (generally a pump) to apply pressure either against an end 106 or end 108 of fuel assemblies 78. The fluid 104 may or may not be able to move past the fuel assembly depending on the specific desired relationship between the wall 110 of the cylinder 100. In any case, the hydraulic fluid 104 may serve in some measure to cool the nuclear fuel assemblies 78. A change in valve setting reverses the flow of hydraulic fluid 104, with a consequent reversal in the direction of the motion of the fuel assemblies 78.

Fig. 13 illustrates another embodiment providing similar operation but through mechanical means. In Fig. 12, a cable or other flexible member 120 is attached between the first ends 122 of a pair of fuel assemblies 78, and a second flexible member 124 is attached between the corresponding second ends 126 of the same fuel assemblies. The fuel assemblies 78 reside in tubes 128. A driving means 126, for example a motor driven pulley, capstan, or spool, engages at least one of the flexible members 120 or 124, and imparts motion to the flexible member, which in turn draws the fuel assemblies 78 between respective first and second positions in reciprocating fashion. Various idler pulleys 132 may be used to ease the passage of the flexible members 120 and 124 through the apparatus.

Claims

What is claimed is:

1. A nuclear power generator comprising, a cylinder, a first nuclear fuel assembly disposed adjacent to said cylinder, a piston reciprocably disposed within said cylinder, a second nuclear fuel assembly supported by said piston for reciprocation with said piston and within said cylinder between a first position wherein said second fuel assembly is spaced from said first fuel assembly and said first fuel assembly is subcritical and a second position wherein said second fuel assembly is disposed adjacent to said first fuel assembly and said first fuel assembly is supercritical, an output shaft, said piston being drivingly connected to said shaft, injecting means for injecting liquid working fluid into said cylinder, and exhaust means for exhausting gaseous working fluid from said cylinder.

2. A nuclear power generator as defined in claim 1 wherein said liquid working fluid is water, and a plurality of high pressure water injectors for injecting water into said cylinder.

3. A nuclear power generator as defined in claim 2 wherein said injectors are spaced around said cylinder and are spaced along the length of said cylinder.

4. A nuclear power generator as defined in claim 1 including turbine means connected to said exhaust means for receiving gaseous working fluid from said exhaust means, said turbine means having an output shaft.

5. A nuclear power generator as defined in claim 4 including condensing means receiving gaseous working fluid from said exhaust means and producing liquid condensate, and means for forcing said condensate adjacent to said first nuclear fuel assembly to cool said first nuclear fuel assembly and to heat said condensate, and further means for forcing said heated condensate to said injecting means.

6. A nuclear power generator comprising a first nuclear reactor having a first nuclear fuel assembly including fresh nuclear fuel, a second nuclear reactor having a second fuel assembly including spent nuclear fuel, said second fuel assembly being disposed adjacent said first fuel assembly, a cylinder disposed adjacent said first and second fuel assemblies, a piston reciprocably disposed within said cylinder, a third nuclear fuel assembly supported by said piston for reciprocation with said piston and within said cylinder between a first position wherein said third nuclear fuel assembly is disposed adjacent to said first nuclear fuel assembly and spaced from said second nuclear fuel assembly so that said first nuclear fuel assembly is supercritical and said second fuel assembly is subcritical and a second position wherein said third nuclear fuel assembly is disposed adjacent to said second nuclear fuel assembly and spaced from said first nuclear fuel assembly so that said first nuclear fuel assembly is subcritical and said second nuclear fuel assembly is supercritical, an output shaft, said piston being drivingly connected to said shaft, injecting means for injecting liquid working fluid into said cylinder, and exhaust means for exhausting gaseous working fluid from said cylinder.

7. A nuclear power generator as defined in claim 6 including a second cylinder disposed adjacent said first and second fuel assemblies, a second piston reciprocably disposed within said second cylinder, said second piston also being connected to said output shaft, a fourth nuclear fuel assembly supported by said second piston for reciprocation with said second piston and within said second cylinder between a first position wherein said fourth nuclear fuel assembly is disposed adjacent to said first nuclear fuel assembly and spaced form said second nuclear fuel assembly so that said first nuclear fuel assembly is supercritical and said second fuel assembly is subcritcal and a second position wherein said fourth nuclear fuel assembly is disposed adjacent to said second nuclear fuel assembly and spaced from said first nuclear fuel assembly so that said first nuclear fuel assembly is subcritical and said second nuclear fuel assembly is supercritical, the first position of said third nuclear fuel assembly occurring concurrently with the second position of said fourth nuclear fuel assembly, and the second position of said third nuclear fuel assembly occurring concurrently with the first position of said fourth nuclear fuel assembly.

8. A nuclear power generator comprising a first nuclear reactor having a first nuclear fuel assembly including fresh nuclear fuel, a second nuclear reactor having a second fuel assembly including spent nuclear fuel, said second nuclear fuel assembly being disposed adjacent said first fuel assembly, a third nuclear fuel assembly supported for movement between a first position wherein said third nuclear fuel assembly is disposed adjacent to said first nuclear fuel assembly and spaced from said second nuclear fuel assembly so that said first nuclear fuel assembly is supercritical and said second fuel assembly is subcritical and a second position wherein said third nuclear fuel assembly is disposed adjacent to said second nuclear fuel assembly and spaced from said first nuclear fuel assembly so that said first nuclear fuel assembly is subcritical and said second nuclear fuel assembly is supercritical, and moving means for moving said third nuclear fuel assembly between said first and second positions.

9. A nuclear power generator as defined in claim 8 wherein said third nuclear fuel assembly when in said first position is disposed within said first nuclear fuel assembly.

10. A nuclear power generator as defined in claim 8 including a tube extending between said first and second nuclear fuel assemblies, said third nuclear fuel being reciprocably disposed within said tube.

11. A nuclear power generator as defined in claim 8 wherein said tube has opposite ends, said moving means comprising a hydraulic circuit connected to said opposite ends.

12. A nuclear power generator as defined in claim 8 wherein moving means comprises mechanical operating means connected to said third nuclear fuel assembly.

13. A nuclear power generator as defined in claim 8 including a fourth nuclear fuel assembly supported for movement between a first position wherein said fourth clear fuel assembly is disposed adjacent to said first nuclear fuel assembly and spaced from said second nuclear fuel assembly so that said first nuclear fuel assembly is supercritical and said second nuclear fuel assembly is subcritical and a second position wherein said fourth nuclear fuel assembly is disposed adjacent to said second nuclear fuel assembly so that said first nuclear fuel assembly is subcritical and said second fuel assembly is supercritical, said moving means also being adapted to move said fourth nuclear fuel assembly between said first and second positions thereof, the first position of said third nuclear fuel assembly occurring concurrently with the second position of said fourth nuclear fuel assembly, and the second position of said third nuclear fuel assembly occurring concurrently with the first position of said fourth nuclear fuel assembly.

14. A nuclear power generator as defined in claim 13 wherein said third nuclear fuel assembly when in its first position is disposed within said first nuclear fuel assembly, and said fourth nuclear fuel assembly when in its second position is disposed within said second nuclear fuel assembly.

15. A nuclear power generator as defined in claim 13 including tubes extending between said first and second nuclear fuel assemblies, said third and fourth nuclear fuel assemblies being disposed within said tubes.

16. A nuclear power generator as defined in claim 15 wherein each of said tubes has opposite ends, said moving means comprising hydraulic connected to said opposite ends.

17. A nuclear power generator as defined in claim 13 wherein said moving means comprises mechanical operating means connected to said third and fourth nuclear fuel assemblies.

18. A nuclear power generator as defined in claim 17 wherein said third nuclear fuel assembly has first and second ends and said fourth nuclear fuel assembly has first and second ends, said mechanical operating means comprising a first flexible member connected between said first end of said third nuclear fuel assembly and said first end of said fourth nuclear fuel assembly, a second flexible member connected between said second end of said third nuclear fuel assembly and said second end of said fourth nuclear fuel assembly, and driving means for driving said cables.

19. The method of generating heat for use in generating power comprising, providing a nuclear power generator comprising a first nuclear reactor having a first nuclear fuel assembly including fresh nuclear fuel, providing a second nuclear reactor having a second fuel assembly including spent nuclear fuel and positioning said second nuclear fuel assembly adjacent said first fuel assembly, providing a third nuclear fuel assembly, and moving said third nuclear fuel assembly in a cycle wherein it is moved from a first position where it is disposed adjacent to said first nuclear fuel assembly and spaced from said second nuclear fuel assembly so that said first nuclear fuel assembly is supercritical and said second fuel assembly is subcritical to a second position wherein said third nuclear fuel assembly is disposed adjacent to said second nuclear fuel assembly and spaced from said first nuclear fuel assembly so that said first nuclear fuel assembly is subcritical and said second nuclear fuel assembly is supercritical, then moving said third nuclear fuel assembly from said second position to said first position to complete a cycle, then repeating said cycle.

20. The method as defined in claim 19 including the steps of providing a fourth nuclear fuel assembly, moving said fourth nuclear fuel assembly in a cycle wherein it is moved from a first position where it is disposed adjacent to said first nuclear fuel assembly and spaced from said second nuclear fuel assembly so that said first nuclear fuel assembly is supercritical and said second fuel assembly is subcritical to a second position wherein said fourth nuclear fuel assembly is disposed adjacent to said second nuclear fuel assembly and spaced from said first nuclear fuel assembly so that said first nuclear fuel assembly is subcritical and said second nuclear fuel assembly is supercritical, then moving said fourth nuclear fuel assembly from its second position to its first position to complete a cycle, then repeating said cycle, the first position of said third nuclear fuel assembly occurring concurrently with the second position of said fourth nuclear fuel assembly, and the second position of said third nuclear fuel assembly occurring concurrently with the first position of said fourth nuclear fuel assembly.