Marcelo O. Giménez. Comisión Nacional de Energía Atómica (CNEA) Centro Atómico Bariloche - Argentina

Transcription

1 CAREM THECNICAL ASPECTS, PROJET AND LICENSING STATUS Marcelo O. Giménez Comisión Nacional de Energía Atómica (CNEA) Centro Atómico Bariloche - Argentina Interregional Workshop on Advanced Nuclear Reactor Technology for Near-Term Deployment, Vienna, 4-8 July

2 CAREM Project Status CAREM is a CNEA (Comisión Nacional de Energía Atómica from Argentina) project to design and build a small nuclear power plant. The first stage is the construction and operation of the demonstration plant (100MWth), CAREM-25, being the base for the development of the commercial versions. Commercial modules power: Natural circulation up to 150 MWe Forced circulation up to 300 MWe 2

3 CAREM Project Status CAREM concept was presented for the first time in a IAEA Conference on Small and Medium Reactors, Perú, In order to enhance safety and to go against the economy of scale (systematically increase of reactor power to reduce costs) and to be an economic option for small and medium electrical grids, CAREM concept is based on: well-proven LWR technology: but re-designing the plant integral PWR design simplicity internalization of Defense-in-Depth since the conceptual design passive safety systems 3

4 CAREM Project Status CAREM design criteria, or similar ones, has been adopted by others plant designers, originating a new generation of reactor designs, of which CAREM was, chronologically, one of the first (Otto Hann) The design basis is supported by the cumulative experience acquired by CNEA+INVAP+Utilities in: Research Reactors design, construction and operation Pressurized Heavy Water Reactors (PHWR) operation, maintenance and improvement. The construction of Atucha II NPP 4

5 CAREM Project Status Evaluated in Generation IV International Forum (USDOE, ), 2002), and selected in the Near Term Deployment group (16 designs selected) After several years of development the CAREM Project reached such a maturity level that the Argentine government decided to support the construction of CAREM demonstration plant (CAREM-25). On December 17 th, 2009, the National Congress declares of interest t the design, construction ti and start-up of CAREM-25. (National Law 26566/2009) 5 5

6 NATURAL CIRCULATION MODULE TECHNICAL DESCRIPTION 6

7 Technical Description Steam Generadores Generators de Vapor Pressurizer Steam Generators Steam Dome Hydraulic control rods drive mechanisms Control rods RCS Pumps RPV Classical loop-type PWR Core Integral-type t PWR 7

8 Technical Description Steam Dome Self pressurization Barrel Hydraulic control mechanisms Steam Generators Natural circulation version Integrated primary cooling system Primary cooling by natural circulation Self-pressurized No Boron in RCS for reactivity control and not required for cold shutdown Reactor Coolant System (RCS) pressure: 12,25 MPa Core outlet, riser and dome temp ~ saturation = 326ºC Core Demonstration plant: 100 MWth: RCS mass flow rate: 410 kg/s 8

9 Temperature distribution along the RCS liquido saturacion Tem mperatura ( o C) Core Riser Steam generators Downcomer Nucleo Chimenea Generadores de Vapor Downcomer Distancia Distance (m) Natural Circulation Modules: operation A change in reactor power implies a change in the mass flow rate, and in the down comer temperature, core outlet temperature remains constant 9

10 Technical Description Reactor Pressure Vessel Diameter: 3,2 m Height: 11 m 10

11 Technical Description Steam Generators 11

12 Technical Description Barrel 12

13 Technical Description Core reflector 13

14 Technical Description Fuel Elements 14

15 Technical Description Absorbing b Elements 15

16 Technical Description Control Rods support Structures 16

17 Technical Description Hydraulic Control Rod Drive 17

18 Technical Description Steam Generators Nozzles 18

19 Technical description Steam Generators 12 identical Mini helical vertical steam generators Once through h type, secondary system in the tube side Secondary pressure: 4,7 MPa Superheated steam: + 30⁰C (290 ⁰ C) Tubes of similar length to equalize pressure loss and superheating Tubes designed to withstand primary pressure (on the outside), without pressure in the secondary side (inside) ( to support a MSLR) 19

20 Technical description Steam to the Turbine Steam Generators feeders Steam Generation Steam Collector SG placed in interspersed positions SG tubes, main steam line and feed water lines, up to isolation valves inside the containment, designed to withstand (inside) the primary pressure for the case of Steam Generator tube rupture event 20

21 Technical description Secondary System Steam Generation SG feed water Secondary System Feed water tank Steam Turbine Condenser 21

22 Technical description Absorbing elements: Fuel Assembly spider : 18 rods Ag In Cd Hexagonal FA with 127 positions: 108 fuel rods Active length: 1,4m 22

23 Technical description Fuel Assembly Burnable poison: Gd 2 O 3 Hexagonal FA with 127 positions: 108 fuel rods 23

24 Technical description Core (Demonstration ti Plant): 61 FFEE U 235 enrichment: 3.1% in all the FFEE, except the one that is in the central position 25 absorbing elements (First Shut down system, reactor subcritical in cold shutdown): 16 to reactivity adjust and control 9 fast shutdown system Fuel cycle: 510 full power days, 50% of core replacement, tailored to customer requirement 24

25 Technical description Hydraulic Control Rods Drive Mechanisms Reactivity it Adjust and Control System (ACS): belongs to the First Shutdown System Cylinder: inlet flow from the Down-comer, movement by steps, controlled by pulses over a base flow no strict requirement on total drop time 25

26 SAFETY APPROACH AND SAFETY SYSTEMS 26

27 CAREM SAFETY ASPECTS Defense in Depth concept is internalized in the design since the conceptual engineering Passive and Simple Safety Systems: neither electricity nor operators actions are required to mitigate accidents during the grace period (for the demonstration plant= 36hrs, each redundancy: x2) 27

28 Technical description Passive Safety Systems SAFETY SYSTEMS PRHRS SSS Two Shutdown Systems: by rods (First SS): FastSS + ACS Boron injection (Second SS) Passive Residual Heat Removal System (PRHRS): condensers Low Pressure Injection System: accumulators 28

29 Technical description Hydraulic y Fast Control Shutdown Rods rods Drives Drive Mechanisms Fast Shutdown System: belongs to the First Shutdown System Cylinder: inlet flow from the Down-comer Piston two positions: top and bottom maximum total drop time: 2s 29

30 Technical description SAFETY SYSTEMS Second Shutdown System (required by the Argentinean Regulatory Body) The SSS is a gravity driven injection device of borated water. It actuates when the Second Reactor Protection System detects the failure of the First SS (ATWS). The system consists of a tank located in the upper part of the containment. It is connected to the reactor vessel by two piping lines: one from the steam dome to the upper part of the tank (pressure equalization line), and the other from the bottom of the tank to the RPV at SG inlet (drainage line). When it is demanded, the valves open automatically and the borated water drains into the RPV by gravity, at reactor pressure. There are two redundancies (1oo2) 30

31 Technical description SAFETY SYSTEMS Passive Residual Heat Removal System (PRHRS) Designed to reduce the RPV pressure and to remove the decay heat. It is a simple and reliable system that operates condensing steam from the RPV in condensers. The condensers have an arrangement of parallel l horizontal U tubes between two common headers. The top one is connected to the RPV steam dome, while the lower header is connected at a position below the RPV water level. (works like an Isolation Condensers) The condensers are located din a pool in the upper part of the containment. The generated steam goes to the suppression pool. System configuration is 2oo4 and pools is 1oo2 (each one 36 hrs) 31

32 Technical description SAFETY SYSTEMS Emergency Injection System The Emergency Injection System prevents core uncovery in case of LOCA. The system consists of one accumulator per redundancy (1oo2) with borated water connected to the RPV. When the pressure in the reactor vessel reaches a relative low value (1.5MPa), the rupture disks break to allow the injection into the RPV. The Passive Residual Heat Removal System is also demanded to help the primary system depressurization, in case of a very small breaks with failure of the SG heat removal. 32

33 Technical description SAFETY SYSTEMS PRHRS and Emergency Injection System designs are based on Functional Reliability Analysis to verify design criteria, using the RMPS methodology ( failure probability of the phenomenology due to uncertainties in engineer and operational parameters and correlations as well) (IAEA Coordination Research Meetings on Development of Methodologies for theassessment of Passive Safety Systems Performance in Advanced Reactors) 33

34 Technical description SAFETY SYSTEMS CONTAINMENT: pressure suppression type, reinforced concrete with stainless steel liner, design pressure 0,5 MPa SSS B PRHRS Pool A Pressure suppression pool 34

35 Technical description SAFETY SYSTEMS CONTAINMENT: FFEE transfer channel 35

36 Technical description SAFETY SYSTEMS CONTAINMENT: SG Feed water headers and main steam collectors Dry-well Wet-well 36

37 Technical description SAFETY SYSTEMS Reactor Protection System: two independent and diverse modules: The First Reactor Protection System demands: First Shutdown System PRHRS containment isolation (ventilation system) )( (LOCA) SG isolation (SGTR) LOCA signal: PRHRS, accumulators (EIS) The Second Reactor Protection System demands: Second Shutdown System 37

38 General strategy to cope with Anticipated Operational Occurrences (AOO) and Design Basis Events (DBE) First: to shut down the reactor (FSS, SSS) Second: to remove the decay heat by means of Defense in Depth (DinD) Level 2 Systems Third: If they fail or there is no AC, the passive safety systems are demanded (DinD Level 3 systems). They lead the plant to a safe state during the grace period (36/72hrs) (the grace period can be extended by simple accident management actions) Once the process systems or the AC is recovered, the control is taken by active systems, final safe state 38

39 SAFETY FUNCTION Power Control Primary System pressure limitation Primary System depressurization and decay power removal Primary System inventory recovery Secondary System pressure limitation Ultimate Heat Sink during the grace period Radio-nuclides confinement Passive SAFETY SYSTEMS (SS-DinD L3) First Shutdown System: Control rods Second Shutdown System: Boron Injection (not for Design Basis Events DBE-) Passive Residual Heat Removal System (PRHRS) Safety Valves (not for DBE) Passive Residual Heat Removal System Emergency Injection System: accumulators Secondary System Safety Valves Pressure suppression and PRHRS pools inside the Containment, no need of external cooling (72 hrs) Pressure suppression-type Passive SS can control and mitigate all the DBE 39

40 Safety related systems Active and Systems other (DinD Defense L2 and in 3 Depth systems Safety Function -Auxiliary SG feed water system L2 and 3 long term-) Residual Heat Removal, as a DinD L2 or after the grace period RPV Inventory reposition, as a DinD L2 or after the grace period Suppression pool, cooling after the grace period Containment pressure reduction, after the grace period -Volume and purification control system: just after SCRAM -RHRS + associated trains to the final heat sink: active systems, from hot to cold shutdown (1oo2) -Volume and purification control system: can cope all LOCAs, (1oo1, but 1oo2 pumps) Suppression pool cooling system: RPV injection mode from the pool (1oo2) Suppression pool cooling system: HXs, 1oo2, active system Suppression pool cooling system: dry-well spray (1oo2), active system No safety grade diesel generators, 1oo2 auxiliary diesels 40

41 SEVERE ACCIDENT PREVENTION AND MITIGATION Despite the very low frequency of core meltdown (Passive Safety Systems+ Grace Period), provision are considered for (DinD Level 4): Severe Accident prevention, grace period extension (under the hypothesis of blackout longer than 72 hrs) by autonomous systems (fire extinguishing external pumps): water injection in the PRHRS pool and PRHRS pool chamber and Containment Suppression Pool cooling Severe Accident mitigation: In vessel Corium retention: RPV external cooling, by gravity Hydrogen passive autocatalytic recombiners 41

42 CAREM: Postulated Initiating Events categorization Category Anticipated Operational Occurrence (AOO) Design Basis Events (DBE) Extended Design Basis Events (EDBE) Occurrence (year 1 ) Acceptance criterion Without core damage, complementing with Anticipated ii ( specific criteria: no RPV safety valve demand, ) DNB&CPR Withoutcore damage, complemented with Unlikely specific criteria: no RPV safety valve demand, ( ) DNBR&CPR>1, no core uncovery (Tclad <Tsat), containment pressure limit. Anticipated Transients Without SCRAM Remote (ATWS, FSS failure), no core damage, ( ) DNBR&CPR>1, RCS pressure limit. (Second Shutdown System + 1 demand of SV) Beyond Design Basis Remote AR acceptability criterion Accidents (BDBA) ( ) (based on Risk evaluation) Severe Accidents (SA) Very remote (<10 6 ) AR acceptability criterion 42

43 Zonas del Primario Domo Zona central Domo Zona periférica Generadores de Vapor NUCLEAR SAFETY Downcomer Núcleo Chimenea ANALYSIS

44 Argentinean Regulatory Acceptability Criterion for NPP (AR ) ARGENTINE ACCEPTANCE CRITERION The objective is to limitit the individual id risk for members of the public associated with potential exposures from a NPP to values not greater than the individual risk associated with normal exposures, commonly incurred by the public in every radiological practice. The individual radiological risk is defined as the probability that, during a certain period and for the nuclear installation, an individual be accidentally exposed to radiation receiving aneffective dose and die due tosuch exposure. Risk (R) is the probability of the intersection of two events: exposure E and fatality F Depends on plant characteristics i (PSA L1,2) and weather conditions (L3) P E P F / E R P E F R P E P F / E Probabilityof a potential Exposure Probability of Fatality due to an Exposure / P F E f D ICRP 44

45 Argentinean Regulatory Acceptability Criterion for NPP Regulationstates that: For accidents with radiological consequences in the public, none accidental group, shall have aean annual probability of occurrence such that, represented graphically according to theeffective effective dose, results in a point located in the non acceptable area of the criterion curve for the public (see Figure). Limit value for the Risk= 10 7 The development of PSA levels 1, 2 and 3 is required 45

46 CAREM SAFETY ASPECTS Characteristics thatenhance safety: preclusion of some classical initiating events (DinD Level 1) Integral primarysystem : very low dose of fast neutrons in the RPV wall (large down comer) No Large Break LOCA No RPV penetrations below the top of SG (good for LOCA and SA) Core cooled by natural circulation: no Loss of Flow Accidents Self pressurized: simplification, nospurious trip of sprays In vessel control and safety rods drives: no rod ejection and LOCA No boron in coolant (operation and shutdown): no boron dilution event A lower number of active components increases plant availability and load factor, reducing the frequency and kind of initiating events. 46

47 CAREM SAFETY ASPECTS Main characteristics that enhance safety given an AOO or DBE: Integral primary system: large thermal inertia simplifies safety evaluations (no loop seal clearance, no SG reflux condensation, no significant pressure waves in case of LOCA hot leg saturated, no stratification in PCS hot leg and surge line, etc.) natural circulation intrinsically enhanced by the lay out Negative reactivity coefficients The Primary System coolant evolves towards saturation in case of Loss of Coolant and in case Loss of Heat Sink due to the PRHRS (behavior easy to model and predict) Self adaptive d i behavior of the RCS mass flow, following the power evolution: enhance safety margins Passive Safety systems and a large grace period 47 47

48 CAREM SAFETY ASPECTS Reactivity insertion: Disparo por alta potencia (108 % - PSPR) Disparo por muy alta potencia (115 % - SSPR) Disparo por alta presión (13 MPa - PSPR) 100 Due to the innovative hydraulic control rods drive, for the Fast 90 Shutdown System Potencia [MW] Límite muy alta potencia (115 % - SSPR) Límite alta potencia (108 % - PSPR) (FSS) and the Adjust and Control System, are Tiempo [s] located inside the RPV, Rod Ejection Accident is avoided. Moreover, as there is no boron in the coolant, boron dilution as reactivity initiating event is precluded. Only inadvertent control rod withdraw transients are possible. Scenarios considering FSS success (DBE) or FSS failure (ATWS) with SSS actuation (EDBE) are postulated: DNBR, CPR and RPV pressure limitsarefulfilledit lfill RCS mass flow rate follows power increment which is an advantage from the safety margins point of view. 48

49 CAREM SAFETY ASPECTS Black Out 330 Loss of Heat Sink/ total loss of SG feed water 320 events 300 Salida de núcleo Salida de chimenea Salida de GV (primario) After reactor shutdown, in the fastest t scenario, 290 the PRHRS Entrada de núcleo is Saturación en domo 280 demanded(i.e.:rcstripparameters)afterhalfanhour,whichgives IPS-SBO Tiempo [s] anidea of the largeinertia of the RCS. Temperatura [ºC] 310 When it is demanded a sharp decrease in pressure is observed till the RCS gets saturated, then pressure decreases smoothly. No safety valvesare demanded (norelief valves). The plant reaches hot shutdown (4.7MPa), by means of the PRHRS, in a safe condition, fulfilling design safety limits, in about 12 hrs. 49

50 CAREM SAFETY ASPECTS 350 Black Out BLACK OUT It isoneof the events with major contribution tib ti 300 to coremeltdown probability in a conventional light water reactor. 290 emperatura [ºC] T Salida de núcleo Salida de chimenea Salida de GV (primario) Entrada de núcleo Saturación en domo In CAREM design it has no impact (it is a DBE) IPS-SBO Tiempo [s] Loss of electrical power produces the interruption of the feed water to the hydraulic CR drives: the absorbing elements drop into the core. The Passive RHRS is demanded due to the initiating event, avoiding the opening of the safety valves. Of course back up diesel generators (1002) are demanded (no required for Safety Systems). Batteries: 36hrs 50

51 CAREM SAFETY ASPECTS 9 8 Loss of Coolant events 4 LOCA 0,0508 m LOCA 0,0381 m 3 LOCA 0,0254 m LOCA00191m 0,0191 Smalldiameter of the penetrations to the RPV, limitedby LOCA 0,0127 m design 2 Apertura Espuria 1 Tope de zona activa de núcleo and come from Safety Systems and containment optimization Large LOCA phenomenology isnot present: ) vel de líquido colapsado (m) Niv tiempo (h) Neither subcooling depressurization nor pressure waves there is not core uncovery, so there is no need of an early water injection to refill the core like in loop type PWR No need of a high pressure injection Safety limits are fulfilled by passive systems: DNBR and CPR limits (this is not the case in classical PWR that use Tclad<1200C, because CHF occurs) 51

52 CAREM SAFETY ASPECTS Steam Generators tube rupture Small loss of coolant to the secondary system (containment by pass) It is mitigated by isolating the affected group of steam generators, closing both the feed water and steam lines (valves located inside the containment and up stream the MSIV). The secondary side of the steam generators (tubes side) reaches thermal equilibrium with the primary circuit, equalising the pressures. End. Eventually the reactor could continue operating at 50% of power. Due to the secondary system flows inside the tubes (lower pressure than in the primary circuit) it)tubes whipping i effect is avoided d and others tubes rupture is not expected 52

53 CAREM SAFETY ASPECTS Main steam line break The sudden depressurisation of the secondary side of the SG increases heat removal from the primary system with the consequent core overpower. Reactor shutdown (FSS or SSS) and Residual Heat Removal System are demanded and the reactor reaches a safe condition, fulfilling DNB and CPR limits. Reactor overpower does not compromise safety limits (DBBR and CPR), also in case of ATWS, because heat removal by the steam generators is intrinsically limited by the reduced water inventory (tube side) in comparison with ihu tubes SG

54 CAREM SAFETY ASPECTS PSA L1 developed to support design and licensing CDF Mitigation by available DinD L2 systems & safety systems Internal initiating events CORE DAMAGE FREQUENCE year -1 E1.A: LOCA S E1.B: LOCA M E2: Loss of power supply to normal busses E3: Generic Transient E4: Loss of SG main feed water E5: Loss of heat sink E6: Steam generator tube rupture E7: Mean steam lines ruptures E8: reactivity insertion E9: Failure of service cooling system 54

55 Facilities and experimental devices to support reactor design 55

56 RA 8 Critical tca facility (at Pilca): neutronic codes validation adato 56

57 Natural Circulation and Self pressurization RIG CAPCN was constructed and operated to produce data in order to study thermo-hydraulic dynamics of the system in conditions similar il to CAREM-25 operational states. (1:1 in height and pressure) 57

58 Natural Circulation and Self pressurization Assessment Many experiments were performed in order to investigate the thermalhydraulic response of the system in conditions similar to CAREM operational states. The influence of different parameters like vapor dome volume, hydraulic resistance and dome nitrogen pressure was studied. Perturbations in the thermal power, heat removal and pressure relief were applied. The dynamic responses at low pressure and temperatures, and with control feedback loops were also studied. It was observed that around the operating point self pressurized natural circulation was very stable, even with important deviation on the relevant parameters. A representative group of transients were selected, in order to check computer models. 58

59 Hydraulic Control Rod Drive Mechanisms Facility CAPEM: High pressure and high temperature rig for testing the innovative Hydraulic Control Rod Drives The construction and start-up at full pressure and temperature finished in 2011 Can be adapted for testing the structural behavior of the FA 59

60 Low Pressure Loop: Hydraulic losses & Flow vibration test 60

61 Thermal Limits and CHF Tests TH LAB IPPE (Obninsk-Russia): LP Freon Loop Test (+250) HP Water Loop Test (25) A facility under development 61

62 Project and Licensing Status 62

63 Project and Licensing Status Documentation equivalent to a Preliminary Safety Analysis Report and the Quality Assurance Manual were presented to the Argentinean Regulatory Body (Federal Authority) at the end of The National Technological University of Avellaneda is performing the Environmental Impact Study (required by Local Authority) 63

64 Project and Licensing Status Facilities at the site are being arranged to start the construction during the first half of 2012 (subjected to a permit by the Argentinean Regulatory Body ). A specific experimental plan will be performed during CAREM-25 preliminary tests and commissioning. Contracts and agreements are under discussion with different Argentinean stakeholders: to perform detail engineering for buildings, containment and process systems for RPV and main components manufacturing 64

65 65 Project and Licensing Status CAREM-25 siting 65

66 Reactor Building (containment inside) Turbine building Auxiliary Building Main access 66

67 THANK YOU! 67