HTGR development in Japan and present status

Transcription

1 HTGR development in Japan and present status Taiju SHIBATA Senior Principal Researcher Group Leader, International Joint Research Group HTGR Hydrogen and Heat Application Research Center Japan Atomic Energy Agency (JAEA) WORKSHOP V VINCO TECHNICAL MEETING 9 th International School on Nuclear Power, 17 th November 2017, Warsaw, Poland

2 Contents 1. Outline of HTTR 2. Graphite components 3. Operation Experience of HTTR 4. Present Status for HTGR development p.1

3 High Temperature Engineering Test Reactor (HTTR) Location of HTTR The Pacific H T T R Japan JAEA 2

4 Bird s eye View of HTTR Reactor Building High Temperature Engineering Test Reactor (HTTR) Fuel handling machine Control room Air cooler Secondary pressurized water cooler Spent fuel storage pool Intermediate heat exchanger Reactor core Primary pressurized water cooler Reactor pressure vessel Reactor containment vessel p.3

5 Major Specifications of HTTR Thermal power Average power density Coolant Outlet coolant temperature Inlet coolant temperature Primary coolant pressure Moderator / Reflector Core height Core diameter Fuel Uranium enrichment Fuel element type Pressure vessel Containment vessel S. Saito et al., JAERI-1332 (1994). 30MW 2.5MW/m 3 Helium gas 850 C/950 C 395 C 4MPa Graphite 2.9m 2.3m Low enriched UO % (Ave. 6%) Prismatic block/ Coated Fuel Particle 2 1/4Cr-1Mo steel 13m(H) 6m(ID) Steel containment 30m(H) 18.5m(ID) p.4

6 High temperature engineering test reactor (HTTR) The first HTGR built in Japan Thermal power: 30MWth Reactor outlet coolant temperature: 950 o C Construction of Reactor Research and development Conceptual design 1969 ~1984 History of HTTR R&D Establishment of fundamental technology ~ First Application and permission of construction Detail design H T T R Basic design System integrity design Conceptual design Construction Experimental very high temperature gas cooled reactor for multipurpose Proposal of a prototype of commercial HTGR s system Start of Loss of Forced cooling test 950 /50 days operation 850 /30 days operation Reactor outlet coolant temperature 950 Safety demonstration test (Control rods withdrawal test) Reactor thermal power (30MW), Reactor outlet coolant temperature 850 criticality Safety demonstration tests Flow rate (%) Power (%) Temp. ( o C) Tripping gas circulators Core cooling flow rate reached zero. Core cooling flow rate Results of test Reactor power Results of test Analysis Maximum fuel temperature Analysis Time (h) No CR reactivity control. No core cooling. Reactor is kept stable naturally with only physical phenomena. Research and development Fuel and Material Oarai gas loop 1 (OGL 1) Reactor physics Very high temperature reactor critical assembly (VHTRC) OECD/NEA project Japan USA France Germany Korea Czech Hungary Thermal Hydraulics A helium engineering demonstration loop (HENDEL) p.5

7 HTTR s design, construction and operational experiments (MHI, Toshiba/IHI, Hitachi, Fuji Electric, KHI and etc.) Design optimization based on extensive technical database Primary coolant system (MHI) Construction of efficient transport and cooling system for very high temperature heat (950 o C) Concentric hot gas duct Primary pressurized water cooler He/He intermediate heat exchanger (IHX) (Toshiba/IHI) Developed new heat (950 o C) resistance material to enable IHX Japanese Technologies for HTTR extraction of heat and making of derivative equipment based on such material HTTR Reactor pressure vessel (Hitachi) Developed new material having high resistance to very high temperature and pressure and construct new pressure vessel using such material Fuel (Nuclear Fuel Industries) Reactor internals (Fuji Electric) Graphite material IG 110 (Toyo Tanso) Reactor internals Advanced technology to coat uranium fuel using ceramics with high radioactivity retaining performance High strength High heat conduction Irradiationresistance p.6

8 Fuel Assembly High density PyC SiC Low density PyC Fuel kernel,.0.6.mm (UO 2 ) 0.92 mm Dowel pin Fuel handling hole 41.mm Fuel channel 8 mm Center hole Graphite matrix Coated fuel particle 39 mm End plug (both sides) Fuel compact Graphite sleeve 580 mm 26 mm Fuel compact 34.mm Fuel rod Dowel socket 360.mm Fuel graphite block Fuel assembly (Pin-in-block type) S. Saito et al., JAERI-1332 (1994). p.7

9 Contents 1. Outline of HTTR 2. Graphite components 3. Operation Experience of HTTR 4. Present Status for HTGR development p.8

10 Graphite in HTTR IG 110 PGX ASR 0RB Bulk density (Mg/m 3 ) Mean tensile strength (MPa)* Mean compressive strength (MPa)* Young s modulus (GPa) (±1/3S u )** Mean thermal expansion coefficient (10 6 K) (293~673K) Thermal conductivity (W/m K) (673K) Ash (ppm) Max. 100 Max Max Grain size (μm) Mean 20 Max. 800 Max * : at room temperature ** : Determined from the cord joining two points (one point is the one-third of the specified minimum tensile strength and the other is the one-third of the specified minimum compressive strength) on the stress-strain curve. JAERI-1332 p.9

11 Graphite Structural Design Code JAERI-1332, JAERI-M Divided into two components from the view point of requirement, replaceability, function and in-service condition Core graphite components Replaceable reflector block (IG-110) Control rod guide block (IG-110) Fuel block (IG-110) Core support graphite components Permanent reflector block (PGX) Hot plenum block (PGX) Support post (IG-110) Lower plenum block (PGX) Carbon block (ASR-0RB) Bottom block (PGX) Requirement : high strength, irradiation stability Replaceability: routine Function : keep fuel at original position, reflector and guide of control rod In-service condition : high neutron fluence (up to n/m 2 ) Requirement : high strength Replaceability: not available Function : support core, reflector In-service condition : low neutron fluence (up to n/m 2 ) Graphite structural design code was established for the HTTR. p.10

12 Integrity Assurance Approach Database Design Manufacturing Inspection Design data including irradiation effect Proof test Structural design code Inspection standards QA/QC management JAERI-M JAERI-M Pre service inspection In service In service inspection method JAERI-Tech JAEA-Data/Code Crack propagation or unexpected oxidation were not supposed First fuel loading Inspection by TV camera Link among Database, Design, Manufacturing and In-service phases are considered. A special committee at Atomic Energy Society in Japan established a draft of standard for graphite core components in HTGR (JAEA-Research ). In service inspection Reactor not in operation Surveillance test Inspection by TV camera Reactor in operation Monitoring of regional temperature distribution p.11

13 Dimensional Change by Irradiation Dimensional change (%) IG-110 HTTR JAEA-Research Irradiation-induced dimensional change of graphite First stage : Shrinkage due to decrease of porosity Second stage: Max. shrinkage (Turn Around: TA) TA depends on irradiation temp. Fast neutron fluence ( n/m 2, E>0.1MeV) Third stage: Expansion from TA due to creation of micro cracks Ref. JAEA-Research p.12

14 Young s Modulus Change by Irradiation Rel. Young s modulus change (E/E 0-1) 照射による縦弾性係数変化率 (E/E 0-1) Irradiation 照射温度 Temperature ( ) ( C) IG Fast Neutron 高速中性子照射量 Fluence( n/m 2, 2 E>29fJ),E>29fJ) Ref.) JAERI-M p.13

15 Thermal Conductivity Change by Irradiation Rel. 照射による熱伝導率の変化率 thermal conductivity change (K/K0) 0 ) Irradiation 照射温度 Temperature ( ) ( ) Fast 高速中性子照射量 Neutron Fluence (10 25 n/m 2,,E>29fJ) E>0.1MeV) IG-110 Ref.) JAERI-M p.14

16 Advantage of IG 110 Graphite Neutron irradiation causes residual stress in graphite components. Safety margin to material data Safety margin to design limit Fig. Example of calculated residual stress for graphite blocks during HTTR operation and limit of stress Ref. J. SUMITA, Characteristics of First Loaded IG-110 Graphite in HTTR Core, JAEA-Technology p.15

17 Contents 1. Outline of HTTR 2. Graphite components 3. Operation Experience of HTTR 4. Present Status for HTGR development p.16

18 High Temperature Continuous Operation Purpose 2010: Reactor outlet helium gas temperature 950 o C 50 days continuous operation To establish fundamental technologies of HTGR To demonstrate stable heat supply to a future heat application system Evaluation of fuel performance (FP retention) Evaluation of core physics Evaluation of impurity control technology in helium coolant Evaluation of IHX performance Evaluation of structural integrity of components Accumulation of operation and maintenance technologies p.17

19 Fuel Performance in HTTR Fractional release of fission gas ( 88 Kr) Operational limit of the HTTR : (U.S.A.) Fort St. Vrain (Germany) AVR (Japan) HTTR Reactor outlet helium gas temperature 950 o C 50 days continuous operation Continuous high temperature operation K. Takamatsu et al., JAEA Research (2010). Operation date in 2010 p.18

20 Loss of Forced Cooling Test Loss of forced cooling (LOFC) & Loss of vessel cooling (LOVC) simulation of station blackout Vessel cooling system (VCS) Vessel cooling system (2 units) Secondary pressurized water cooler (SPWC) Core is cooled from the outside by radiation and natural convection. Water cooling tube Reactor RPV Heat Auxiliary cooling system Intermediate heat exchanger (IHX) Primary pressurized water cooler (PPWC) Heat removal adjustment panel Cooling water : Stop of circulator and pump p.19

21 Loss of Forced Cooling Test Loss of forced cooling test (LOCF test) : Stop of all circulators in primary circuit Stop of all primary circulators flow rate of primary coolant : 100% 0%) Test Date Test Condition Dec.21, 2010 Increase of reactor core temperature Decrease of reactor power due to negative reactivity feedback ( resonance absorption of neutron in U 238 ) Reactor power and fuel temperature remain in stable condition Ref ) Leaflet of HTTR, JAEA. Flow rate (%) Power (%) Temp. ( o C) Initial reactor power 30% (9MW) Without scram (no movement of CR) Test Results Stop of circulator Core flow rate Test result Reactor power Test result Peak fuel temperature Analytical result Elapsed time (hr) 20 p.20

22 Contents 1. Outline of HTTR 2. Graphite components 3. Operation Experience of HTTR 4. Present Status for HTGR development p.21

23 Policies of HTGR Development in Japan Technical development of HTGR is stated in the following policies approved by the Cabinet. Strategic Energy Plan approved by the Cabinet on April 11, 2014 Under international cooperation, government of Japan facilitates R&D of nuclear technologies that serve the safety improvement of nuclear use, such as hightemperature gas cooled reactors which are expected to be utilized in various industries including hydrogen production and which has inherent safety. Growth Strategy 2017 approved by the Cabinet on June 9, calls for future R&D concerning the HTGR development to be promoted using JAEA s HTGR test reactor and through international cooperation. Strategic Roadmap of hydrogen and fuel cell issued by the committee in the METI on June 23, METI : Ministry of Economy, Trade and Industry p.22

24 Ongoing Activities under MEXT MEXT established a committee including MEXT, METI, JAEA, industries and universities to discuss roadmap and conceptual design for the first demonstration plant. Specification of commercial HTGR, R&D plan, introduction scenario are being discussed. Industry Vendors Users (Electricity/Hydrogen/Heat Utilization) Toshiba Corporation Nippon Steel & Sumitomo Metal Corporation Hitachi, Ltd. Iwatani Corporation Fuji Electric Co., Ltd. Chiyoda Corporation Mitsubishi Heavy Industries, Ltd. Toyo Engineering Corporation Fuel/Graphite manufactures JGC Corporation Nuclear Fuel Industries, Ltd. Hitachi Zosen Corporation Toyo Tanso Co., Ltd. Trading company/think tank Toyota Motor Corporation Marubeni Utility Services, Ltd. Nissan Motor Co., Ltd. Canon Institute for Global Studies Honda R&D Co.,Ltd. Academy Government University of Tokyo Ministry of Education, Culture, Sports, Science and Tokyo Institute of Technology Technology (MEXT) Tokyo City University Japan Atomic Energy Agency (JAEA) Toyo University of Agriculture and Technology Kyushu University Observer: Japan Electrical Manufacturers Association, Japan Atomic Power Company, Institute of Applied Energy, Ministry of Economy, Trade and Industry (METI) p.23

25 Overview of HTGR and Heat Application Technologies HTTR (1) Reactor technology Reactor outlet coolant temperature 950 o C at 30 MWt (April 2004) 950 o C / 50 days operation (March 2010) Advanced fuel development HTTR tests for HTGR safety enhancement Safety review by NRA is underway (3) Commercial HTGR design Design study of commercial HTGR systems Core design of plutonium burning HTGR (2) Heat application technologies He compressor Hydrogen facility Completion of basic technologies related to hydrogen production facility and gas turbine power generation Establishment of operation control technology and facility reliability for IS process 31 hrs. hydrogen production with 0.02m 3 /h (October 2016) (4) HTTR GT/H2 test Coupling to HTTR Licensing demonstration Plant performance test Establishment of safety design philosophy and international standardization for commercial HTGRs Integrated demonstration of HTGR heat application system technologies p.24

26 HTTR GT/H 2 Test (Basic Design Outline) Objective of HTTR GT/H 2 test Demonstration of system technologies for HTGR helium gas turbine power generation and H 2 production Establishment of safety standard and design consideration for coupling between reactor and heat application system Improvement of cost evaluation reliability Recuperator / Precooler Secondary heat exchanger Helium gas turbine Plant cycle schematic Reactor IHX Isolation valves Gas turbine Precooler H 2 facility Generator Recuperator HTTR Isolation valves Reactor Intermediate heat exchanger (IHX) HTTR GT/H 2 test facility (Planned) p.25

27 Thank you for your attention. JAEA is willing to cooperate future HTGR program with Japanese mature technologies! Taiju SHIBATA

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