Method for managing stoppage of a pressurised-water nuclear reactor

Abstract

Disclosed is a method for managing stoppage of a pressurized-water nuclear reactor integrated into a submerged module for producing electrical power, in case of detection of a primary/secondary leak in a steam generator equipped with a safety valve, which generator is connected to the reactor and associated with a standby cooling unit. The method includes: detecting a primary/secondary leak in the steam generator; automatically stopping the reactor and isolating the broken steam generator; bringing the corresponding standby cooling unit online; monitoring the primary pressure and, once the primary pressure has passed below the set pressure of the safety valves of the steam generators, isolating the standby cooling unit of the broken steam generator; and continuing to passively cool the reactor with the remaining steam generators and cooling unit.

Claims

1. A method for managing stoppage of a submerged pressurized-water nuclear reactor of a module configured to produce electrical power, the module having a plurality of steam generators each being provided with a safety valve and associated with a standby condenser, the method comprising: detecting a primary/secondary leak of primary fluid into secondary fluid of one of the steam generators; automatically stopping the reactor and isolating the steam generator suffering from the primary/secondary leak; bringing the standby condensers online; monitoring a primary pressure of the primary fluid; determining that the primary pressure is below the set pressure of the safety valve of the steam generator, then isolating the standby condenser of the steam generator suffering from the primary/secondary leak by preventing the primary fluid from being supplied to the standby condenser; and continuing to passively cool the reactor with remaining steam generators of the plurality of steam generators and condensers.

2. The method for managing stoppage of the submerged pressurized-water nuclear reactor according to claim 1, wherein the detection of the primary/secondary leak is done by detecting one or several of the following phenomena: increased activity due to contamination by the primary fluid into the secondary fluid, increased secondary water inventory, and decreased primary water inventory.

3. The method for managing stoppage of the submerged pressurized-water nuclear reactor according to claim 1, wherein the isolation of the standby condenser of the steam generator suffering from the primary/secondary leak is obtained by triggering a controlled valve inserted between the standby condenser and the steam generator suffering from the primary/secondary leak.

4. The method for managing stoppage of the submerged pressurized-water nuclear reactor according to claim 3, wherein the controlled valve is associated with a compressed air supply.

5. The method for managing stoppage of the submerged pressurized-water nuclear reactor according to claim 2, wherein the isolation of the standby condenser of the steam generator suffering from the primary/secondary leak is obtained by triggering a controlled valve inserted between the standby condenser and the steam generator suffering from the primary/secondary leak.

Description

BRIEF DESCRIPTION OF THE DRAWINGS

(1) The invention will be better understood upon reading the following description, provided solely as an example and done in reference to the appended drawings, in which:

(2) FIG. 1 is a flowchart illustrating the different steps of a method according to the invention,

(3) FIG. 2 is a block diagram illustrating part of a pressurized water reactor, and

(4) FIG. 3 shows a graph illustrating the mitigation of the primary/secondary leak.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS

(5) The fundamental issue in managing a primary/secondary leak on a pressurized-water reactor with passive and automated safety with no atmospheric discharge lies in detecting the damage and obtaining a compromise between cooling and mitigation of the leak.

(6) The solution proposed in the present application therefore consists of: a primary/secondary leak detection logic using the conjunction of several signals respectively based on an increase in the N-16 activity of the secondary circuit, an inconsistency in the regulation of the supply water maintaining the level of the steam generator, an inconsistency of the regulation of the primary volumetric control ensuring maintenance of the primary water inventory, an automation stopping cooling by the damaged steam generator.

(7) A primary/secondary leak is characterized by: increased activity of the secondary circuit due to contamination by the primary fluid, increased secondary water inventory, decreased primary water inventory.

(8) In the present application, we propose to detect this leak from one or several of these phenomena.

(9) Indeed, one could settle for detecting nitrogen 16, which could suffice in itself, but in the case of the present application, we for example choose to consider that there is a primary/secondary leak in the case of a simultaneous presence of all three of the previously indicated criteria, in order to guarantee the absence of untimely detection.

(10) It should in fact be recalled that in the case of the present FLEXBLUE project, a non-detection of the leak does not call the nuclear safety into question, but leads to significant deterioration of the industrial tool, this being the problem that the present application proposes to resolve.

(11) The first criterion above, i.e., the increased activity of the secondary circuit, is detectable by using nitrogen 16 detection systems, placed on the primary steam lines of the reactor.

(12) These detection system includes sensors that must be associated with a triggering threshold that is variable based on the power extracted by the reactor making it possible to distinguish an increase in activity due to functional and consecutive leaks and an increase in power of the core, from an actual primary/secondary leak.

(13) The second criterion is detectable via an inconsistency on the system for regulating the level of the steam generator.

(14) Indeed, this system is based on the conjunction of two input values, i.e., the power extracted by the turbine with which a tabulated steam flow rate is associated and the reading of the level of steam generator.

(15) From these data, the regulating system injects the liquid supply water flow rate (ARE) necessary to keep the level of the steam generator constant.

(16) Thus, the unexpected increase in the secondary water inventory will be detectable by a difference between the expected supply water flow rate for the supplied power and the supply flow rate actually injected to keep the level of the steam generator constant.

(17) The last criterion is detectable via an inconsistency on the volumetric regulating system of the primary circuit.

(18) If the latter must inject a large quantity of water to keep the level of the pressurizer at a constant primary pressure and temperature, there is an unidentified loss of primary inventory.

(19) The objective of the proposed automation is therefore to find a compromise between: ensuring maximum power discharge and rapid cooling of the primary while using all of the standby condensers (or other cooling means), and the need to isolate the cooling of the damaged generator quickly so as not to feed the leak and to limit the loss of primary inventory.

(20) All of this is of course done while keeping a completely passive action mode, like all of the standby systems of the project in progress, i.e., according to category D of the reference classification established by the AIEA.

(21) To that end, the applicant has conducted simulations, and various phenomena have been observed.

(22) Indeed, if the standby cooling means of the damaged steam generator are stopped too soon, there is an increase in the pressure and the water level in this steam generator, causing the overpressure valves of this steam generator to open and leading to contamination of part of the confinement enclosure.

(23) If the standby cooling means of the damaged steam generator are stopped too late, there is an excessive loss of primary inventory, causing automatic depressurization and flooding of the confinement enclosure.

(24) The invention proposes to carry out the method as illustrated in FIG. 1, which leads to ensuring the safety of the assembly when stopped.

(25) In this case, it is of course necessary to refer to the various documents filed by the applicant on the FLEXBLUE project to obtain a complete description of the reactor.

(26) The described method leads to stopping the standby cooling means of the damaged generator by basing itself on an indication reflecting the energy inventory of the primary circuit so as to be sure that the power to be discharged at the primary, which can only decrease, does not cause valves to open.

(27) To that end, the method illustrated in FIG. 1 is implemented.

(28) This method can be summarized as follows:

(29) If a primary/secondary leak has been detected, the standby cooling of the damaged steam generator is isolated once the thermohydraulic conditions of the boiler prove that there is no longer any risk of opening valves of this steam generator if its cooling is interrupted.

(30) In the case described in the present application, one decides to isolate the standby condenser of the damaged steam generator if the primary pressure falls below a value lower than the set value of the protection valves of the steam generator.

(31) Indeed, the method illustrated in FIG. 1 then includes a step 1 for detecting a primary/secondary leak that, in step 2, triggers an automatic stoppage of the reactor and an isolation of the steam generator.

(32) In step 3, the standby condensers are brought online, and in step 4, the primary pressure is read so that it is compared to the set pressure of the valves of the steam generators.

(33) When the latter, i.e., the primary pressure, becomes lower than the set pressure of the valves, the isolation of the standby condenser of the damaged steam generator is triggered in step 5, then in step 6, the passive cooling of the reactor continues on the remaining exchangers.

(34) The implementation of this method can be illustrated by the means described in FIG. 2.

(35) To provide a simplified description of what has already been developed by the applicant, we will simply note that this FIG. 2 shows a submerged module 9, a pressurized-water reactor vessel designated by general reference 10, a steam generator designated by general reference 11 and a standby condenser designated by general reference 12, for example in contact with an endless cold source 13, such as the ocean or the like, as described in the various documents relative to the FLEXBLUE project.

(36) For safety reasons, valves 14 and 15 used to bring the standby condensers, like the condenser 12, online, are lined and are of the type that is normally open such that a loss of logistic support of the module automatically causes the commissioning of the standby condensers.

(37) The sequence proposed in the present application must counter this phenomenon.

(38) Thus, an isolating valve designated by general reference 16 is provided in this connecting line to the standby condenser, to isolate the latter under the action of a control valve designated by general reference 17, connected to a compressed air source 18, for example.

(39) Thus, when regulating air is lost following the loss of electrical power supplies following the stoppage of the reactor, the standby condenser 12 is isolated by the valve 16 placed downstream, supplied via the dedicated air supply 18 and for example positioned by a valve 17 of the pyrotechnic type or any other device having the same qualities, i.e., an extremely low call failure rate, and very energy-efficient maneuvering and maintenance in position.

(40) Of course, different technical solutions to the problem of automatically isolating the secondary cooling can be considered.

(41) It is also possible to consider an entire series of alternatives on two specific technical points of the described elements, i.e., the logics associated with the three detection criteria to trigger the automations and the thermohydraulic criteria used to quantify the minimal primary energy loss before isolating the standby condenser from the damaged steam generator.

(42) Lastly, it is possible to consider applying the same isolating logics of the standby cooling systems on active safety reactors.

(43) The sequence can be simpler, with the understanding that this is limited to permanent isolation of the atmosphere discharge valve of the damaged generator once the primary thermohydraulic conditions allow it (i.e., a low enough residual power, a primary pressure lower than the set pressure of the protection valves of the steam generator, etc.).

(44) FIG. 3 for example shows a simulation of the operation of the entire mitigation sequence in the context of the project.

(45) The computing code used is the ATHLET code, which is a thermohydraulic simulation tool used in nuclear accident studies and recognized by international safety authorities.

(46) The obtained results show the operation covering the entire range of primary/secondary leaks commonly required during a safety demonstration.

(47) For a guillotine-type break of a steam generator tube, the results shown in FIG. 3 are obtained, reflecting quick management of the damage no longer using only passive systems after the detection and not causing unacceptable or irreversible damage to the industrial tool.

(48) In this FIG. 3, the beginning of the leak is designated by reference 20, the detection of the damage by reference 21 causing the automatic stoppage of the reactor and the supply of steam to the turbine.

(49) Reference 22 shows the startup of the standby condensers, and 23 shows the stoppage of the standby condenser of the damaged steam generator.

(50) Reference 24 shows the primary/secondary equilibrium achieved when the leak is eliminated, the reactor next cooling slowly by its passive means.

(51) These curves are established from the primary pressure, the pressure of a first healthy steam generator and a second steam generator affected by the leak.

(52) Of course, this model is simplified, in particular in terms of the pressure and steam generator level regulations.

(53) However, it provides a good overview of the obtained results.

(54) This also makes it possible to increase and improve the arguments put forth regarding the intrinsic and passive safety of nuclear reactors, as defined by the applicant.

(55) Of course, still other embodiments can be considered.

(56) In particular, the invention could of course be suitable for a pressurized-water reactor integrated into a land module.