The Pebble Bed Modular Reactor Design and Technology Features

Transcription

1 The Pebble Bed Modular Reactor Design and Technology Features Frederik Reitsma Frederik Reitsma Independent Representative South Africa Advanced Nuclear Reactor Technology for Near Term Nuclear Safety Management Deployment Latest Developments in Germany and South Africa Wednesday, 15 October 2008, University of the Witswatersrand, IAEA Johannesburg, Interregional Workshop 4-8 July

2 Energy required for development IAEA Interregional Workshop 4-8 July

3 Preamble - Status of PBMR Company PBMR (Pty) Ltd, is a South African engineering company that was dedicated to the design, licensing and realisation of the Pebble Bed Modular Reactor. The project to build a demonstration unit was abandoned in 2010 and all the employees were formally retrenched on 30 September The PBMR company still exist as an entity and according to a government decision it will be maintained till at least Its new role is to Care and Maintain the developed Intellectual Property. PBMR appointed a team of 9 engineering specialist on contract who with support staff must fulfil this role. The strategy and practical implementation of it, is under developed. Currently the Test facilities are in Care and Maintenance Fuel development laboratory, Helium test facility (HTF) Heat Transfer Test Facility at North West University IAEA Interregional Workshop 4-8 July

4 Presentation layout Introduction and Background The Pebble Bed Modular Reactor PBMR400 design The Safety concept Plant configurations for alternative applications Implications from the Fukushima event Test Facilities Concluding remarks IAEA Interregional Workshop 4-8 July

5 INTRODUCTION AND BACKGROUND Why HTGCR s Historical basis The pebble bed reactor concept IAEA Interregional Workshop 4-8 July

6 Why High-temperature Gas-cooled Reactors? Significantly improved safety High temperatures lead to higher efficiency than conventional nuclear plants Attractive economics (small, modular, to be proven in future) Market is growing for smaller reactors Smaller reactors lend themselves to distributed generation (advantages relate to grid stability and transmission costs) Extended scope of application due to higher temperature availability; process heat / supply of process steam for petro-chemical industry future hydrogen production Pebble Bed Reactors offers enhanced non proliferation characteristics. The use of Th-232 in a HTR (specifically an on-load fuelling Pebble Bed Reactor) with U-233 recycle could significantly reduce reliance on uranium resources and reduce lifetime of spent fuel waste IAEA Interregional Workshop 4-8 July

7 Historical GCR Background Early Gas Cooled Reactor (GCR) Development 3.5 MWt X-10 open-circuit air cooled reactor at Oak Ridge (1943) 2 MWt Saclay unit in France (1951), initially nitrogen cooled, then carbon dioxide Commercial GCR Plants* Calder Hall in the U.K. (1953) Magnox Plants {U.K.(26), France (8), Japan, Italy and Spain (1 each)} Advanced Gas Cooled Reactor Plants, {U.K.(14)} *These commercial GCRs are graphite moderated, carbon dioxide cooled IAEA Interregional Workshop 4-8 July

8 Initial HTGR Development HTGRs include ceramic coated fuel particles with a graphite moderator and helium cooling. Prototype Plants Dragon Reactor Experiment 20MWt, 750 C Core outlet, U.K./OECD project, (first power July 1965) Peach Bottom (No.1) 115MWt/40MWe, 725 C Core outlet, U.S.A., (critical march 1966), Total generation = 1.36 billion kwh. IAEA Interregional Workshop 4-8 July

9 HTR fuel elements Block and pebble IAEA Interregional Workshop 4-8 July

10 Where the pebble bed idea was born 1944 / USA Secret report by F Daniels Suggestions for a High-Temperature Pebble Pile IAEA Interregional Workshop 4-8 July

11 and where did the coated particle idea came from 1960 / UK R.A.U Huddle patent Triso coated particle IAEA Interregional Workshop 4-8 July

12 Concept by Prof-Dr Rudolf Schulten, at the research centre at Jülich Conceptualized and developed since the 1950 s Technology is built on a wealth of know-how developed in Germany over 40 years. Arbeitsgemeinshaft Versuchsreaktor 46MWt / 15MWe, 850 / 950 C Core outlet, first electricity, 1967, closed 1988 Total generation = 1.67 billion kwh. AVR operated 21 years Safety illustrations The German Legacy Many different pebble fuel types tested HTR-Module (80 MWe) Modular design Inherent safety concept certification IAEA Interregional Workshop 4-8 July

13 Pebble bed demonstration plants Thorium High Temperature Reactor (THTR-300) 750MWt/300MWe, Germany, Pebble bed core, a helium cooled core consisting of a HEU-Th fuel cycle, the primary system enclosed in a pre-stressed concrete reactor vessel. Steam conditions of 530 C/530 C. First power in 1985, closed 1989 Total generation ~ 2.9 billion kwh IAEA Interregional Workshop 4-8 July

14 The pebble bed reactor concept A pebble bed core is a loosely packed bed of billiard sized spherical fuel elements Fuel spheres are added from the top and extracted at the bottom Fuel can be recycled through the core a number of times to flatten the power profile by achieving a more homogeneous and flatter burnup profile Burnup is measured when fuel is extracted at the bottom spent fuel (above a selected value) is discarded and a fresh fuel element is loaded in its place usable fuel is returned for another pass through the core Fuel load rate is a function of the number of passes and keeping the reactor critical (burnup) Provide flexibility in changing fuel loading and even fuel types IAEA Interregional Workshop 4-8 July

15 THE PBMR-400 PEBBLE BED REACTOR Main characteristics Primary circuit components Reactor core Fuel design IAEA Interregional Workshop 4-8 July

16 What is the PBMR-400 IAEA Interregional Workshop 4-8 July

17 Reactor Characteristics PBMR400 Pebble bed modular reactor Fuel as LEU UO2 coated particles in 60mm diameter spheres Online refueling Annular core configuration allows higher thermal power but still maintain the inherent safety features of the modular design concept Long slender core Needed for inherent safety heat removal Annular core geometry provides for short heat transfer path to the outside of RPV resulting in lowering of maximum fuel temperature during a loss of cooling event Thickness of fuel annulus restricted Shutdown systems only in reflector (practical considerations / THTR experience) RPV maximum diameter manufacturing cost considerations (single supplier) Central reflector Restricts maximum fuel temperatures during DLOFC Allow secondary shutdown system to be diverse and far removed from the control rods also shutdown with single system Online refueling and Multi-pass Low excess reactivity (load only the fuel you need) No need for shutdown to re-fuel Flatter burnup profile with increased number of passes IAEA Interregional Workshop 4-8 July

18 Reactor Main Design Data Rated Power per Module 400 MW(th) Refueling 6 times; 3 load / de-fuel chutes Inlet / outlet helium T 500 / 900 C System pressure 90 bar Enrichment 9.6% Startup enrichment ~4.2% Discharge burnup ~92 GWd/te Coolant bypass flow ~20% - influence temperatures Fuel residence time ~930 days Fresh fuel load per day ~486 Max power per fuel sphere ~2.8 kw/fs Max power density ~11 MW/m 3 Max fuel T (normal operation) ~1070 C Max DLOFC temperature ~1580 C IAEA Interregional Workshop 4-8 July

19 PBMR400 Features Passive safety characteristics achieved by inherent design features design rules out core melt all ceramics fuel Inherent safety features proven during public tests System shuts itself down Can show that there is very limited need for safety grade backup systems Helium coolant is chemically inert (single phase) Coated particle provides excellent containment for the fission product activity Large thermal capacity lead to slow thermal transients No common mode failure in the core (a single fuel failure do not lead to additional failures) Ingress of water into core eliminated by design and air ingress limited No need or reduced off-site emergency plans (smaller safety zone) Low proliferation risk On-load refueling Distributed generation due to smaller size Modularity Lower capital costs High efficiency (> 41%) IAEA Interregional Workshop 4-8 July

20 Brayton power conversion cycle helium-cooled, direct-cycle Net Electrical Efficiency ~41% IAEA Interregional Workshop 4-8 July

21 PBMR-400 Power Conversion Unit Components IAEA Interregional Workshop 4-8 July

22 Building Layout IAEA Interregional Workshop 4-8 July

23 Reactor Layout Main Characteristics: annular core of 3.7 m fixed central reflector of 2 m, an effective cylindrical core height of 11 m, a graphite side reflector of ~90 cm, 24 partial length control rod positions in the side reflector as the reactivity control system (RCS) eight Small Absorber Sphere (SAS) systems positioned in the fixed central reflector as the Reserve Shutdown System (RSS) filled with 1 cm diameter absorber spheres containing B4C when required, three fuel loading positions and three fuel unloading tubes core contains ~ 452,000 fuel spheres uranium loading is 9 g per fuel-sphere U235 enrichment at 9.6 wt%. IAEA Interregional Workshop 4-8 July

24 Reactor Layout Fuel Line Top Reflector Fuel Core Centre Reflector Control Rod Side Reflector SAS Channel Bottom Reflector SAS Extraction Point IAEA Interregional Workshop 4-8 July

25 Core characteristics - Power 0 o bottom) [cm] Axial height (top t Channel 1 Channel 2 Axial height [cm m] Power per fuel sphere [kw] Flow channel Pass 1 Pass 2 Pass 3 Pass 4 Pass 5 Pass Channel 3 Channel Channel Relative pow er density IAEA Interregional Workshop 4-8 July

26 Core characteristics Fuel temperature 40% 35% Average fuel sphere temperature Average fuel (UO2) temperature (Doppler) Maximum fuel (UO2) temperature (kernel in centre) 38.6% Percentage of pebbles at temperature 30% 25% 20% 15% 10% 5% 4.6% 5.1% 2.9% 3.4% 2.8% 2.5% 4.8% 6.5% 13.4% 15.5% 0% 0.0% 0.0% Temperatures (C) IAEA Interregional Workshop 4-8 July

27 Core characteristics Burnup Axial height (top to bottom) [cm] channels; 1st pass 5 channels; 2nd pass 5 channels; 3rd pass 5 channels; 4th pass 5 channels; 5th pass 5 channels; 6th pass Burnup [MWd/t] IAEA Interregional Workshop 4-8 July

28 Core characteristics Fluxes / Spectrum 3.5E+14 Fluxes [n cm m-2 s-1] 3.0E E E E+14 > 0.1 MeV 0.1 MeV - 29 ev 29 ev ev < 1.86 ev 1.0E E E Radial distance [cm] IAEA Interregional Workshop 4-8 July

29 x Thermal neutron flux [cm-2.s-1] z in cm r in cm Figure 1: Thermal flux (<1.86 ev) profile IAEA Interregional Workshop 4-8 July

30 Core characteristics - Fuel temperature 0 Temperatures [C] Distance from top of core [c cm] Flow Channel 1 Flow Channel 2 Flow Channel 3 Flow Channel 4 Flow Channel IAEA Interregional Workshop 4-8 July

31 Pebble bed multi-pass principle IAEA Interregional Workshop 4-8 July

32 THE FUEL IAEA Interregional Workshop 4-8 July

33 The fuel design Fuel spheres: Units Values Pebble radius cm 3.0 Thickness of fuel free zone cm 0.5 Density of graphite in matrix/fuel free zone g/cm U-235 enrichment of uranium wt% 9.6 Coated particles: Kernel diameter µm 500 Kernel density g/cm Coating material C / C / SiC / C Layer thickness µm 95 / 40 / 35 / 40 Layer densities g/cm / 1.90 / 3.18 / 1.90 IAEA Interregional Workshop 4-8 July

34 Triso-coated particle 1 mm IAEA Interregional Workshop 4-8 July

35 AUXILIARY SYSTEMS IAEA Interregional Workshop 4-8 July

36 Reactor Cavity Cooling System Remove heat from the reactor cavity Normal operation as well as decay heat Maintain RPV and concrete temperatures Water cooled Active and passive system design / modes Several days grace time including boil off time before water makeup needed Ultimate heat sink is the concrete structure if the RCCS fails RCCS important for investment protection IAEA Interregional Workshop 4-8 July

37 Example of detailed CFD analysis of the RCCS Concrete RCCS Air Temperature RPV IAEA Interregional Workshop 4-8 July

38 Fuel Handling and Storage System On-load refuelling Reduction of required excess reactivity Dynamic core movement ensures that temperature profile through fuel spheres changes continually and no amoeba effect is possible Fuel Handling System similar to designs used in THTR Design can incorporate spent fuel storage tanks onsite for the life of the plant IAEA Interregional Workshop 4-8 July

39 THE PBMR-400 SAFETY CONCEPT Safety of NPP revisited The modular pebble bed philosophy Inherent characteristics and passive safety PBMR-400 fulfilling the fundamental safety principles IAEA Interregional Workshop 4-8 July

40 Changing thoughts Safety revisited Comments made during the opening session of PHYSOR2010 Safety is no longer the major public concern the handling of the waste are. Nuclear growth path is steep, very steep. Is there a place, time and resources for another technology? What does it mean for modular HTGR Is the price we pay for inherent safety too large? requires a power density of about 1/10 th of PWR s HTGRs can also provide process heat electricity only ~30% of total energy needs is this the opportunity we need? Today the Safety of Nuclear Power Plants designs are once again under scrutiny IAEA Interregional Workshop 4-8 July

41 HTR-Modul (200 MWth) Germany Modular concept Siemens, 1985 Basic safety principles: - Passive decay heat removal - Shutdown by rods in side reflector - No intolerable reactivity excursions - rod expulsions - water ingress - Steam generator below core - Coated particle the sole containment IAEA Interregional Workshop 4-8 July

42 HTR-Modul principle DIMENSIONS AND POWER ARE FIXED BY INHERENT PROPERTIES [can not be chosen as usually] Diameter: given by shutdown from outside D ~ 300 cm Power density: given by maximum fuel temperature [T = 1600 o C] Q ~ 20 MW/m Core height: given by blower [dp~ 1.5 bar, Xenon] H ~ 10 m This yields a maximum power per Modul of: P max = MW th IAEA Interregional Workshop 4-8 July

43 INHERENT CHARACTERISTICS AND PASSIVE SAFETY IAEA Interregional Workshop 4-8 July

44 Engineered SAFETY or the preferred approach, Inherent Safety. IAEA Interregional Workshop 4-8 July

45 Inherent safety philosophy Safety does not rely on engineered systems that may fail but on the inherent design and the laws of physics. Increased safety: smaller plant that allows for reactor cooling by passive heat transfer mechanisms following an accident prevents the fuel temperatures from increasing to levels where significant radioactive fission products can be released from the fuel and thus eventually into the atmosphere, the type of accident that is most feared by the public. IAEA Interregional Workshop 4-8 July

46 Design for Nuclear Safety must demonstrate for PBMR-400: Fundamental Safety Functions Reactivity Control Heat Removal Confine Radioactivity IAEA Interregional Workshop 4-8 July

47 Reactivity Control Inherent safety design characteristics Large negative temperature feedback effects Automatic shutdown with loss of coolant Strong negative temperature coefficient limits reactivity excursions Doppler -3.2x10-5 ρ/ C; Total -3.8x10-5 ρ/ C at operating temperatures Low excess reactivity On-line reloading Excess reactivity only to overcome power changes / load follow Reactivity Design functionalities Reactor shutdown during operation and maintaining sub-criticality for cold conditions Reactivity control during operation and daily load-follow Damped xenon oscillations, also for all operator actions Inherently safe features during operation and licensing events Fuel storage sub-criticality for fresh, used and spent fuel IAEA Interregional Workshop 4-8 July

48 Reactivity Control Negative temperature coefficients of reactivity Temperature coefficient of reactivity 1.0E E E E E-04 Fuel Moderator Reflector Total E-04 Temperatures (due to iso-deltic changes applied) [C] IAEA Interregional Workshop 4-8 July

49 Reactivity Control Automatic reactor shutdown Safety demonstration: Stop coolant flow and no control rod movements Power (% %) Total Power (%) Fission Power (%) Average Fuel Temperature Average Moderator Temperature Temperature e (C) Time (seconds) IAEA Interregional Workshop 4-8 July

50 Reactivity Control Automatic power reduction Safety demonstration at HTR-10 reactor during HTR conference in Beijing, during HTR2004 conference, September 2004 IAEA Interregional Workshop 4-8 July

51 Heat Removal Inherent safety design characteristics Passive heat removal post-shutdown decay heat removal is achievable through conduction, natural convection and radiation heat transfer, due to the core geometry, low power density of the core and high thermal capacity of the core structures Needs low power density! Centre Reflector Pebble Bed Side Reflector Core Barrel RPV RCCS Citadel Conduction Radiation Conduction Conduction Radiation Convection Convection Conduction Radiation Conduction Radiation Convection Convection Convection Conduction Radiation IAEA Interregional Workshop 4-8 July

52 Heat Removal Passive heat removal * Heat removal under all reactor operation conditions and events * Two active heat removal systems: - Self-sustained PCU-Brayton thermodynamic cycle - Core Conditioning System (CCS) used during maintenance * Passive heat transfer from the core to the outer heat sink during loss of forced cooling * CCS designed as defense-in-depth to keep the core at normal operating temperatures during upset conditions * RCCS heat removal or concrete and the surrounding earth as the ultimate heat sink Illustration of Fuel Temperatures behaviour for a DLOFC Event over 60 days IAEA Interregional Workshop 4-8 July

53 Confine Radioactivity Fuel containing radio-active fission products Adequate Confinement of Radioactivity is ensured by: - High-quality ceramic coated-particle fuel of proven design - Sufficient Heat Removal IAEA Interregional Workshop 4-8 July

54 Confine Radioactivity Fuel containing radio-active fission products Fuel elements with multi-coated fuel particles are used for optimum retention of fission products The silicon carbide layer has the ability to contain fission products Can withstand very high temperatures. IAEA Interregional Workshop 4-8 July

55 Confine Radioactivity Additional barriers to fission products and radio-active active releases * Transport of radioactivity through two main mechanisms: - Neutrons and gammas originating at the reactor and activation - Radio-nuclides in the coolant having escaped from the fuel spheres Barriers: The pressure boundary, building, suppression pool, filters, etc IAEA Interregional Workshop 4-8 July

56 Confine Radioactivity Prevention of massive corrosion Air ingress remains an important subject in PBMR safety analysis Prevent air corrosion of graphite by providing robust reactor isolation and limiting air supply All the different break location has been analysed Engineering solutions such as pipe support design can minimize the break gap and reduces air ingress Inert gas injection systems are effective in stopping air ingress Water ingress limited in direct cycle power conversion system IAEA Interregional Workshop 4-8 July

57 Safety Functions - Recall The most important inherent characteristics of the PBMR which contribute to the fulfillment of the fundamental safety functions are: A fuel and core design with a low excess reactivity and an overall negative temperature coefficient of reactivity sufficient to accommodate any foreseeable reactivity insertions during start-up and power operations without damage to the fuel. A core design that ensures that post-shutdown decay heat removal is achievable through conduction, natural convection and radiation heat transfer, due to the core dimensions, low power density of the core and high thermal capacitance of the core structures. Peak temperatures remain below the structural design limits, and the fuel temperature is kept below the limit where serious degradation of the coated particles would lead to a significant activity release. High-quality ceramic coated-particle fuel of proven design, which adequately retains its ability to confine radioactive fission products over the full range of operating and accident conditions. IAEA Interregional Workshop 4-8 July

58 Four Principles of Stability Incorporated into the PBMR Design Core may never melt or be overheated to unallowable temperature Thermal stability Nuclear transients may never lead to unallowable power output excursions or cause unallowable fuel element overheating Fuel elements may never be allowed to corrode excessively Nuclear stability Chemical stability Reactor cannot melt, practically no release of fission products, catastrophe-free nuclear energy Core may never be allowed to deform or change composition Mechanical stability IAEA Interregional Workshop 4-8 July

59 PLANT CONFIGURATION FOR DIFFERENT APPLICATIONS IAEA Interregional Workshop 4-8 July

60 Standardised Nuclear Heat Supply System IAEA Interregional Workshop 4-8 July

61 Electricity plus low temperature steam for desalination 700 ⁰C 570 ⁰C η ± 40% IAEA Interregional Workshop 4-8 July

62 Cogeneration plant - electricity plus process steam 750 ºC 570 ºC IAEA Interregional Workshop 4-8 July

63 Different quality process steam supply High P Process Steam Extraction Medium P Process Steam PBMR Reactor Heat Exchanger/ HP Steam Steam Helium ~ Generator Steam Turbine-Generator 750 C ROT Waste Heat or Low P Steam MW Other House Loads Water Treatment Power Sales Reclaimed Water Blowdown Potable water IAEA Interregional Workshop 4-8 July

64 Electricity Plant IAEA Interregional Workshop 4-8 July

65 LESSONS LEARNED AND DESIGN IMPLICATIONS FROM THE FUKUSHIMA EVENT IAEA Interregional Workshop 4-8 July

66 Implications from the Fukushima event Since the South African project has been discontinued no design evaluation performed after Fukushima Only two aspects to be highlighted The specific characteristics of a packed pebble bed of fuel spheres due to an earthquake Impact on the South African nuclear future IAEA Interregional Workshop 4-8 July

67 EARTHQUAKE PEBBLE BED COMPACTION IAEA Interregional Workshop 4-8 July

68 Pebble Beds and earthquakes The impact of earthquakes on the PBMR design was investigated as part of the safety case Shaker-table experiments (SAMSON) located at the HRG (Hochtemperatur-Reaktorbau GmbH) site at Jülich, Germany used to postulate conservative compaction densities and times for use in the safety studies Focus of the safety evaluation: compaction of the pebble-bed or fuel region only no radial disturbance in the core cavity dimensions - excluded by the core structure and graphite reflector design change in the bulk or average packing density during an earthquake study core-neutronics and thermal-hydraulics behaviour of a postulated SSE IAEA Interregional Workshop 4-8 July

69 SAMSON Facility SAMSON experiments at 0.4 g > (5 seconds) > (15 seconds) IAEA Interregional Workshop 4-8 July

70 PBMR400 SSE postulated event Postulate: Only effect is pebble bed compaction Decrease in pebble bed or core effective height Very conservative assumptions for concept design Packing fraction increases: i) > 0.62 ii) > 0.64 No control rod movement Compaction duration: i) 5 seconds ii) 15 seconds Longer duration of strong shaking will not compact the core beyond the assumed values Includes a PLOFC and DLOFC (beyond design base) Reactivity increase due to: Denser packing of fuel spheres Reduction of control rod effectiveness IAEA Interregional Workshop 4-8 July

71 Phenomena and restrictions The two major phenomena: neutronic response of the fuel due to the bed compaction (streaming, leakage, spectrum changes, temperature feedback) changes in the heat transfer (pebble bed packing fraction, reduced core height) quantify the changes in: the core reactivity fission power material temperatures fuel heat-up rate during the power excursion IAEA Interregional Workshop 4-8 July

72 Fission power (SSE + PLOFC) (1 st very conservative results showing no cliff edge effects) IAEA Interregional Workshop 4-8 July

73 Core average fuel sphere temperatures (SSE + PLOFC) IAEA Interregional Workshop 4-8 July

74 Actual SSE results: SSE Fission Power as a % of the Steady State Values (0 s to 30 s) with RPS Trip Initiated on the Reactor Power Control rod insertion begins at 1.73 s as a result of the power SCRAM set point Power (%) Time (s) IAEA Interregional Workshop 4-8 July

75 FUKUSHIMA: IMPLICATIONS ON THE SOUTH AFRICAN NUCLEAR OUTLOOK IAEA Interregional Workshop 4-8 July

76 Future of Nuclear in South Africa (IRP2) The cabinet approved the country's 20-year Integrated Resource Plan on 16 March 2011 the new plan foresees 23% of all new plants coming on stream between now and 2030 to be nuclear nuclear would supply MW generally accepted to be only LWR / PWR technology (Generation II+/III) The government said earlier that it would not put its planned nuclear expansion on hold, despite concerns over nuclear safety given the Fukushima event. Energy minister Dipuo Peters has said the South Africa government is aware of the risks of nuclear power and that the government will factor this into account when it maps out a detailed proposal for new nuclear construction. IAEA Interregional Workshop 4-8 July

77 TEST FACILITIES AND R&D PBMR Micro Model Helium Test Facility Heat Transfer Test Facility Plant Control Room / Training Simulator Fuel Manufacturing and Development IAEA Interregional Workshop 4-8 July

78 PBMR Micro Model - NWU o o The Pebble Bed Micro Model (PBMM) demonstrated the operation of a closed, three shaft, pre- and inter-cooled Brayton cycle with a recuperator. Construction started in Jan 2002 and commissioning was completed on 23 Sep o o Also used for code verification and validation Included in the IAEA CRP-5 TecDoc V&V IAEA Interregional Workshop 4-8 July

79 Helium Test Facility (HTF) Nov 2004: Construction of the Helium Test Facility (HTF) commences at Pelindaba, 1st tests in Jan Design Nov Apr 2006 Construction May Oct 2006 Commissioning Nov 2006 Phase 1Acceptance tests Dec 2006-Mar 2007: Training of PBMR and operation under ISTN supervision Apr 2007-Mar 2010: Independent plant operation and testing by PBMR Apr 2010 current: Mothballed HTF is a non-nuclear facilitythat tests full scale systems and components at PBMR Demonstration Power Plant design conditions. Height: 40 m Levels: 8 Foot print: 130 m 2 Crane capacity: 20 ton IAEA Interregional Workshop 4-8 July

80 Helium Test Facility Located at Pelindaba IAEA Interregional Workshop 4-8 July

81 HTF Simplified Process Flow Diagram Legend:Actual tested/used operating envelope (max design limits not yet tested) FHS 9 MPa 550 C (700 C) RCS 280 C RSS 9 MPa 550 C 640 C (900 C) HICBS Main Loop 9 MPa 280 C 14 MPa 9 MPa Blower 9 MPa 450 C IAEA Interregional Workshop 4-8 July

82 Reactivity Control System Stepper motor scram test setup Secondary shock absorber drop tests Prototype Control Rod Drive Mechanism chain test setup Prototype Control Rod Drive Mechanism to be tested in HTF Functionality successfully tested IAEA Interregional Workshop 4-8 July

83 Reserve Shutdown System Full scale discharge vessel tests in HTF SAS Storage Container SAS Transport Pipe Valve actuator rod to be tested in HTF Graphite Structures SAS Outlet Assembly Functionality successfully tested IAEA Interregional Workshop 4-8 July

84 HTF Fuel Handling System (FHS) Sphere Indexers Gas Brakes Helium Return Core Loading Device > successful sphere passes completed The functioning of the sphere counter was tested Leak tests were done on shaft penetrations, to determine the helium leak flow rate Sphere Lines Test Vessel Helium Supply Valve Blocks Core Unloading Device IAEA Interregional Workshop 4-8 July

85 Heat Transfer Test Facility North West University High Temperature Test Unit (HTTU) o Pressure 1 bar o Core Temperature 1600 ºC o Nitrogen & Helium High Pressure Test Unit (HPTU) Pressure 50 bar Temperature 100 ºC Nitrogen IAEA Interregional Workshop 4-8 July

86 Heat Transfer Test Facility Heat Transfer Test Facility (HTTF) consists of two test units located at NWU: High Pressure Test Unit (HPTU): used to perform pressure drop tests, braiding effects tests and convection coefficient tests under high pressure but at ambient temperatures. High Temperature Test Unit (HTTU): used to perform natural and forced convection tests and velocity profile tests under high temperature but at atmospheric pressure. Sep 2005: Construction commences at North West University and 1st tests in Aug Last planned tests were completed in Oct IAEA Interregional Workshop 4-8 July

87 HTTF Status Heat Transfer Test Facility Status High Temperature Test Unit (HTTU) High Pressure Test Unit (HPTU) Status Complete Natural Convection Header Comparison Tests Pressure Drop Tests (0.36 ; 0.39 ; 0.45) Complete Complete Velocity Profile Header Development Convection Coefficient Tests (0.36 ; 0.39 ; 0.45) Complete Complete Forced Convection Header Tests Near-Wall Test Section Complete Complete Vacuum Header Braiding Effect Tests (0.36 ; 0.39 ; 0.45, Random beds) Complete Note: Comparison tests to further understanding of differences between annular and cylindrical cores with respect to modern CFD code development. Comparison Tests Small Cylindrical Randomly Packed Bed Test Small Annular Randomly Packed Bed Test Complete Complete IAEA Interregional Workshop 4-8 July

88 HTTF (HPTU + HTTU) Summary 3 Years of operation by 12 Test engineers Accident free manhours 18 Test Reports 150 Technical Memos Operating Experience HPTU Test results demonstrate the applicability of the relevant KTA correlations for pebble beds. Improved understanding of Random vs Structured bed behavior and simulation thereof. HTTU New Model for Effective Thermal Conductivity Major graphite degradation was experienced at high temperatures. This was problem was successfully resolved, allowing for completion of tests program with no further outages IAEA Interregional Workshop 4-8 July

89 Plant Control Room Training Simulator IAEA Interregional Workshop 4-8 July

90 DPP 200 Indirect (Rankine) Cycle Main System Overview IAEA Interregional Workshop 4-8 July

91 Operator Training Simulators DPP400 Plant Simulator Development of the DPP400 direct cycle simulator was started in 2001 to get familiarization with the direct cycle HTR simulation requirements and the various platforms available. Simulator Hardware: Computers and typical control room equipment. Software: From GSE and ABB, using C++, Fortran and VB The ABB control software is used to control plant models with the soft controllers emulating the control hardware. DPP200 Plant Simulator Indirect cycle plant Models based on those developed for the DPP400. The reactor model scaled down to 200MWt. IAEA Interregional Workshop 4-8 July

92 Fuel manufacturing / fuel development Mar 2007: Advanced Coater Facility commissioned (5kg). Mar 2008: Necsa apply for construction and cold commissioning license for Fuel Plant. Dec 2008: The Fuel Development Laboratories, based at Pelindaba successfully manufactured coated particles. Jan 2009: Coated particles containing 9.6% enriched uranium shipped to the USA for irradiation testing at the Idaho National Laboratory. Sept 2009: The first High Temperature Reactor fuel spheres or pebbles containing 9.6% enriched uranium is manufactured. Sixteen of these spheres were earmarked for irradiation tests to demonstrate the fuel s integrity under reactor conditions (tests were not performed). Many achievements Understanding some FP transport mechanisms (silver), Leach test results Advanced fuel studies IAEA Interregional Workshop 4-8 July

93 Kernel Manufacturing IAEA Interregional Workshop 4-8 July

94 Fuel Fabrication Kernel Casting IAEA Interregional Workshop 4-8 July

95 CONCLUDING COMMENTS Generation IV characteristics Conclusion IAEA Interregional Workshop 4-8 July

96 Generation IV Goals Sustainability 1.Generate energy sustainably, and promote long-term availability of nuclear fuel 2.Minimize nuclear waste and reduce the long term stewardship burden Safety & Reliability 3.Excel in safety and reliability 4.Have a very low likelihood and degree of reactor core damage 5.Eliminate the need for offsite emergency response Economics 6.Have a life cycle cost advantage over other energy sources 7.Have a level of financial risk comparable to other energy projects Proliferation Resistance & Physical Protection 8.Be a very unattractive route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism IAEA Interregional Workshop 4-8 July

97 The Pebble Bed Modular Reactor: Progress towards a Generation IV candidate. Sustainability Studies on U, Pu and Th cycles Increase burnup - > 80,000 MWd/te already proven -> much higher potential CARBOWASTE program, reprocessing technologies Graphite matrix has excellent FP and Actinide containment properties Safety & Reliability (the walk away safe principle, no immediate action needed) Inherent safety properties, large heat capacity Reliability to be proven but many issues already resolved from the prototypes No core damage in traditional sense for modular design Graphite corrosion limitation important No off-site impact already shown by current designs Economics Challenge due to low power density no free lunch Fuel manufacturing process to be streamlined, graphite recovery Availability factor can be excellent (on-line refueling, magnetic bearings etc.) Small modular (short construction time, large volumes (production line approach) Small incremental capital cost; reduce losses in long transmission lines Need a demonstration plant to prove this Proliferation Resistance & Physical Protection Kernels SiC coating can be broken (a lot of hard work) Unattractive Pu ratio s even after 1-pass IAEA Interregional Workshop 4-8 July

98 Conclusions Modular Pebble Bed Reactors has advanced safety characteristics passive safety achieved by the inherent design characteristics. The design displays many of the characteristics of future nuclear plants (Generation IV reactors) The Modular Pebble Bed Reactors can participate in the total energy market and not only in electricity generation The technology is built on a wealth of know-how developed in Germany over 40 years and recent enhancements made in South Africa, China and USA (NGNP). The pebble bed community now wait with great anticipation for the HTR-PM to take the technology into the main stream NPP offering. IAEA Interregional Workshop 4-8 July

99 The Pebble Bed Modular Reactor Design and Technology Features Frederik Reitsma THANK YOU FOR YOUR ATTENTION Advanced Nuclear Reactor Technology for Near Term Nuclear Safety Management Deployment Latest Developments in Germany and South Africa Wednesday, 15 October 2008, University of the Witswatersrand, IAEA Johannesburg, Interregional Workshop 4-8 July