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US 9,852,820 B2Granted
Nuclear steam supply system
AssigneeSmr Inventec, LLC
InventorsKrishna P. Singh, Joseph Rajkumar
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AssigneeSmr Inventec, LLC
InventorsKrishna P. Singh, Joseph Rajkumar
Estimating the term…
A nuclear steam supply system utilizing gravity-driven natural circulation for primary coolant flow through a fluidly interconnected reactor vessel and a steam generating vessel. In one embodiment, the steam generating vessel includes a plurality of vertically stacked heat exchangers operable to convert a secondary coolant from a saturated liquid to superheated steam by utilizing heat gained by the primary coolant from a nuclear fuel core in the reactor vessel. The secondary coolant may be working fluid associated with a Rankine power cycle turbine-generator set in some embodiments. The steam generating vessel and reactor vessel may each be comprised of vertically elongated shells, which in one embodiment are arranged in lateral adjacent relationship. In one embodiment, the reactor vessel and steam generating vessel are physically discrete self-supporting structures which may be physically located in the same containment vessel.
A nuclear steam supply system with natural gravity-driven coolant circulation, the system comprising:
A nuclear steam supply system with natural gravity-driven coolant circulation, the system comprising:
A nuclear steam supply system with natural gravity-driven coolant circulation, the system comprising:
Dependencies are inferred from the claim text.
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From the US filing history. Foreign family members aren’t listed.
Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants: promoting flow of the coolant
Promoting flow of the coolant: by convection, e.g. using chimneys, using divergent channels
Details of nuclear power plant: primary side of steam generators
Energy generation of nuclear origin
About 7 years 2 months of term left · ends 12 Dec 2033 (estimated)
The present invention relates nuclear steam supply systems, and more particularly to a steam supply system for a small modular reactor having a natural gravity driven coolant flow circulation system.
Pressurized water reactors (PWRs) for nuclear power generation facilities utilize both pumped and natural circulation of the primary coolant to both cool the reactor core and heat the secondary coolant to produce steam which may be working fluid for a Rankine power generation cycle. The existing natural circulation PWRs suffer from the drawback that the heat exchange equipment is integrated with and located within the reactor pressure vessel. Such an arrangement not only makes the heat exchange equipment difficult to repair and/or service, but also subjects the equipment to corrosive conditions and results in increased complexity and a potential increase in the number of penetrations into the reactor pressure vessel. In addition, locating the heat exchange equipment within the reactor pressure vessel creates problems with respect to radiation levels encountered for crews to repair the heat exchange equipment in proximity to the radioactively hot components of the reactor vessel. The general view has also been that the heat exchangers should be located in the reactor vessel to achieve natural circulation in those systems which may utilize this type of flow circulation.
An improved nuclear steam supply system is desired.
The features of the exemplary embodiments of the present invention will be described with reference to the following drawings, where like elements are labeled similarly, and in which:
FIG. 1 is cross section of a containment vessel including a nuclear steam supply system according to the present disclosure which includes a steam generator assembly comprising a reactor vessel and a steam generating vessel;
FIG. 2 is a schematic flow diagram of the nuclear steam supply system of FIG. 1 showing the flow path or circulation flow loop of primary coolant represented by the flow arrows;
All drawings are schematic and not necessarily to scale.
The features and benefits of the invention are illustrated and described herein by reference to exemplary embodiments. This description of exemplary embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,”, “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivative thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Accordingly, the disclosure expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features.
Referring to FIGS. 1-6, a steam supply system for a nuclear pressurized water reactor (PWR) according to the present disclosure is shown. From the thermal-hydraulic standpoint, the system includes a steam generator assembly 100 generally including a reactor vessel 200 and a steam generating vessel 300 fluidly coupled to the reactor vessel. The steam generating vessel and reactor vessel are vertically elongated and separate components which hydraulically are closely coupled, but discrete vessels in themselves that are thermally isolated except for the exchange of primary loop coolant (i.e. reactor coolant) flowing between the vessels. As further described herein, the steam generating vessel 300 in one embodiment includes a preheater 320, main steam generator 330, and a superheater 350 which converts a fluid such as water flowing in a secondary coolant loop from a liquid entering the steam generating vessel 300 at inlet 301 to superheated steam leaving the steam generating vessel at outlet 302. The secondary coolant loop water may be a Rankine cycle fluid used to drive a turbine-generator set for producing electric power in some embodiments.