Status of HTTR Project in JAEA

Transcription

1 Status of HTTR Project in JAEA Kazuhiko Kunitomi Nuclear Hydrogen and Heat Application Research Center Japan Atomic Energy Agency (JAEA) TWGGCR meeting at IAEA March 5, 2013

2 The HTTR Project HTTR (1) Reactor Technology Completion of continuous 50 -day high temperature (950 o C) operation in March 2010 Safety demonstration test Various R&D on innovative HTGR (2) Hydrogen Production Technology HTTR IS Process Material development Efficiency improvement Integration technology First achievement of continuous H 2 production by IS process in the world (3) Innovative HTGR Plant Design Naturally safe HTGR (NSHTR) based on fully inherent features Clean Burn HTGR (CBHTR) for surplus plutonium burning Multi-Purpose Small-sized HTGR (MPSHTR) for emerging nations such as Kazakhstan GTHTR 300 series for Middle East 1

3 Outline of HTTR HTTR Graphite-moderated and helium-cooled VHTR Intermediate heat exchanger (IHX) Fuel Rods Graphite Block Reactor pressure vessel Major specification Thermal power 30 MW Fuel Coated fuel particle / Prismatic block type Core material Graphite Coolant Helium Inlet temperature 395 C Outlet temperature 950 C (Max.) Pressure 4 MPa First criticality : 1998 Full power operation : o C operation 2004 Containment vessel Hot- gas duct 2

4 Safety Demonstration Test using the HTTR (1/4) Purposes To demonstrate safety in accident conditions To convince regulator and public of safety of the HTGR Past Tests Reactivity insertion (FY 02-06) Partial loss of primary forced cooling (FY 02-06) Complete loss of primary forced cooling (FY 10) Stop of all circulators (simulation of loss of forced cooling) : 30% of full power Stop of all circulators + stop of one out of two units of the vessel cooling system : 30% of full power Future Tests Complete loss of primary forced cooling Stop of all circulators (simulation of loss of forced cooling) : 80%, 100% of full power Stop of all circulators (simulation of all blackout) + stop of all units of the vessel cooling system : 30% of full power OECD/NEA LOFC Project (3 years, until March ) The OECD/NEA established a Loss of Forced Cooling (LOFC ) project using HTTR for better understandings of HTGR safety(the tests shown in red color are classified as test s for OECD/NEA LOFC projects). Participants: USA, Korea, Czech, France, Germany, Hungary, Japan 3

5 Safety Demonstration Test using the HTTR (2/4) Temp. ( o C) Flow rate (%) Power (%) Loss of forced cooling test (LOCF test) : Stop of all circulators in primary circuit Stop of all primary circulators flowrate 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 Initial reactor power 30% (9MW) Without scram (no movement of CR) Test Results 0 Stop of circulator Core flow rate Test result Reactor power Test result Peak fuel temperature Analytical result Elapsed time (hr) 4 4

6 Safety Demonstration Test using the HTTR (4/4) Loss of forced cooling (LOFC) & Loss of vessel cooling (LOVC) simulation of station blackout HTTR Flow Schematic Vessel cooling system (2 units) SCS Secondary pressurized water cooler (SPWC) Vessel cooling system (VCS) Core is cooled from the outside by radiation and natural convection. Upper panel Upper radiation shielding Side panel Radiation shielding Thermal reflector plate Water cooling tube ACS Auxiliary heat exchanger CV Reactor Intermediate heat exchanger (IHX) PCS Primary pressurized water cooler (PPWC) RPV Heat Heat removal adjustment panel Coolin g water : Stop of circulator and pump Lower panel 5

7 Naturally Safe HTGR (NSHTR)(1/2) The most important consideration is understanding of safety by public. Probabilistic Safety Analysis (PSA) The best way for nuclear experts but not for average people. Average people cannot understand how much safe of the system with the probability of the core melting of 10-6 /ry, 10-8 /ry, etc. The easiest way for public Demonstrable safety Most of the worst accidents are simulated by real reactors and safety is confirmed. For the accident condition that the simulation test cannot be conducted by real reactors Safety shown by analysis The safety in the worst accident postulated reasonably is shown by analysis. What is the worst accident postulated reasonably? The cause of the accident is technically or scientifically postulated. Example Cracks of RPV is postulated. Explosion of RPV is not postulated. Proposal of Naturally Safe HTGR (NSHTR) 6

8 Naturally safe HTGR (NSHTR)(2/2) Naturally-safe HTGR (NSHTR) ensures the reactor safety by natural phenomena not by safety systems. Evaluation of natural phenomena is not an issue of PRA analysis. Evaluation of natural phenomena is performed by deterministic analysis. Natural phenomena prevents the large amount of FP release Confinement function Reactor Fuel coating Phenomena which degrades confinement function Diffusive release Sublimation Fission product Uranium Corrosion Failure Causes of events Core heat up Oxidation Inherent Safety Feature (natural phenomena) Doppler effect conduction, Radiation, Natural convection Oxide layer formation Attain stable state Retain radionuclides within confinement RPV CV Fuel coating Flammable gas explosion Flammable gas oxidation Safety is proven by deterministic way and not by probability for public understanding. 7

9 Clean burn HTGR (CBHTR) Incineration of surplus plutonium from LWR spent fuel Basic concept Employ Inert Matrix Fuel (IMF) for fuel kernel covered with ceramic layers. Inert Matrix Fuel(IMF) is composed of PuO2 and Yttria Stabilized Zirconia (YSZ), NOT composed of MOX. Fuel Compact Fuel Block Fuel Rod Kernel(PuO 2 -YSZ) Buffer Layer ipyc Layer SiC Layer opyc Layer Coated Fuel Particle Combination of IMF and CFP technology developed by JAEA is the key of this concept 8

10 Clean burn HTGR (CBHTR) Advantage of the CBHTR High nuclear proliferation resistance due to IMF fuel kernel Plutonium cannot be extracted easily. JAEA has developed various technology for IMF fuel since 1990s. High plutonium incineration ability No plutonium is generated from the kernel because no U-238 is in the kernel High fuel integrity during the operation Combination of IMF fuel and confinement function of TRISO prevents FP release from the coated fuel particle in the high burn-up condition. Reactor Type Pu-239 Consumption(%) Fuel Burn-up (GWd/t) LWR-MOX 55 Pu-U MOX 50 CBHTR 95 PuO 2 -YSZ 550 High durability of spent fuel in geological repository Combination of IMF fuel and confinement function of TRISO keeps FP inside the coated fuel particle stably. 9

11 Multi-purpose Small-sized HTGR(MPSHTR) Collaboration with National Nuclear Center in RK Multi purpose HTGR at Kurchatov in Republic of Kazakhstan Hydrogen production (670 ) (850 ) Power generation by gas turbine (GT) 620 IHX: Intermediate Heat Exchanger SG : Steam Generator IS : Iodine sulfur process GT : Gas turbine ST : Steam turbine MWt or 30 MWt , 12MPa MPa District heat Power generation by steam turbine (ST) 10

12 Development of IS process H 2 (m 3 /h) Electric heater Glass Electric heater Glass Electric heater Industrial materials Nuclear demonstration Lab-scale test (~1997) Bench scale test (~2004) Process engineering test (~2010) HTTR-IS test Verify the closed-cycle H 2 production The reactants except water are fully regenerated and reused with the process. Hydrogen (H 2 ) 450 HTGR 900 Oxygen (O 2 ) A 100l/h-scale test apparatus is under construction. Partial tests have been started. Production of hydrogen iodide (HI) and sulfuric acid (H 2 SO 4 ) Oxygen production unit H 2 + I2 2HI Production of HI and H 2 SO 4 H 2 SO 4 1/2O 2 + SO 2 +H 2 O 2HI + H 2 SO 4 Iodine (I) Sulfur (S) cycle I 2 + SO 2 + 2H 2 O cycle SO 2 I 2 + H 2 O H 2 O Hydrogen iodide (HI) decomposition Water Sulfuric acid (H 2 SO 4 ) decomposition Hydrogen production unit Completion of one-week continuous hydrogen production with automatic control system ref) S. Kubo et al., Proc. GLOBAL 2005, Tsukuba, Japan, Oct. 9-13, 2005, No

13 Future plan of HTTR project Innovative HTGR Plant design Design of Naturally-safe HTGR (NSHTR) Design of Clean Burn HTGR (CBHTR) Design of Multi-Purpose Small-sized HTGR (MPSHTR) R&D for NSHTR and CBHTR (fuel, graphite, gas turbine, etc.) Commercial HTGR System Power generation by steam turbine and process steam supply in 2020s Commercial HTGR/VHTR System Reactor Technology (HTTR) Long-term (50 days) high-temp. operation to demonstrate reactor characteristic (2010) Safety demonstration test including Loss-of- Forced Cooling test, etc. Demonstration test for MPSHTR, NSHTR and CBHTR Hydrogen Production Technology IS Process 1 week continuous hydrogen production (2004) Process engineering test (A new test apparatus is under construction from 2010.) Hydrogen Production with HTTR-IS System (max.1000m 3 /h) NSHTR and CBHTR for commercial use in 2030s Hydrogen production for commercial use in 2030s GTHTR300C 12

14 Procedure of re-start of HTTR HTTR facility Earthquake : March 11, 2011 Buildings Important structures for the seismic design Confirmation test at cold condition (at 4MPa) Check of the leakage Check of the pressure drop Check of the vibration Completed (from May to June 2011) 2011 Sept. 20 Safety report was required by MEXT Detailed check Check of the integrity of the concrete surface by detailed visual examination Check of integrity of the component support structures Seismic analysis submitted on Sept. 7, 2012 Evaluation of the structural integrity using measured seismic waves Time history response analysis of reactor buildings, structures and etc. New safety standards for research rectors (will be issued on Dec Repair of wall surface and Periodical inspection Final evaluation for the reactor start-up by NRA Re-Start Agreement of the reactor startup with local governments 13

15 Core differential 炉心差圧 pressure [kpa] (kpa) Safety check under low temperature condition Confirmation of plant conditions by non-nuclear heating (coolant temperature: 120 o C, pressure: 4MPa) 〇 primary coolant leakage %/day(before the earthquake: 0.24 %/day, limit: 0.28%/day) 〇 Gas circulators, CRDM, RSS normal 〇 Plant data No change (showing core differential pressure as an example) 2010 平成 (before 22 年 the earthquake) 3 台停止試験前試験の日付 12/1 12/2 12/3 12/4 12/5 12/6 12/7 12/8 12/9 12/10 12/ 次 HGC (before 炉心差圧 the earthquake) 今回 次 HGC (after 炉心差圧 the earthquake) 前回 :00 昇温開始 Heatup :00 昇圧開始 Pressure increased 10:00 Reached 4MPa 到達 4MPa 17:00 降圧開始 Pressure decreased 0.0 5/20 5/21 5/22 5/23 5/24 5/25 5/26 5/27 5/28 5/29 5/30 平成 (after 年 the earthquake) 地震後確認試験の日付 No degradations were observed in components after the earthquake 14

16 Tentative Schedule of the HTTR FY 2012 FY 2013 FY 2014 Integrity confirmation Detailed check Seismic analysis Safety Report Submission (Sept. 7) Regulatory, etc. Evaluation by NRA Negotiation with local gov. New safety standards for research rectors HTTR facility Repair and periodic inspection Repair of concrete surface Periodic inspection (after evaluation by NRA) Operation Re-start Safety demonstration tests (LOFC test, etc.) 15

17 Summary We propose a Naturally Safe HTGR (NSHTR) and Clean Burn HTGR (CBHTR) in response to the Fukushima Daiichi LWR accident. The safety of the NSHTR is assured by natural phenomena. No engineered safety features, AC power, or prompt actions by plant personnel are required even in the worst accident. For public acceptance, we propose a deterministic approach instead of the probabilistic approach. The CBHTR efficiently incinerates surplus plutonium from LWR. Preparation for the process engineering test of the IS process is on going. Inspection and seismic evaluation of the HTTR have been completed. Evaluation by NRA is underway. 16

18 Reference

19 Demonstration and R&D items for NSHTR Demonstration test using HTTR All black-out All CRs withdrawal Rapid and large coolant flow increase Confirmation of low radioactive material inventory in primary circuit R&D on Fuel and Graphite Experimental evaluation of oxidation of SiC layer of coated fuel particles Development of oxidation resistant graphite (ex. surface C/SiC layer) Fuel rod Confirmation of low radioactive material inventory Graphite sleeve hold fuel compacts and make coolant flow channels. Fuel compact Oxidation of fuel sleeve May cause mechanical damage on coated fuel particles (CFP) 18

20 R&D items for CBHTR R&D on Fuels for CBHTR Investigation on fuel manufacture process and quality control techniques Irradiation Test Development of analytical code for fuel design Investigation on erosion-corrosion of TRISO fuel in repository 19

21 Safety for Earthquake and Blackout - GTHTR300 - Position (m) Earthquake Shutdown Reactor scram Blackout Control rods insertion AC power is unnecessary Air Cooling by natural convection Heat Exchanger vessel Control valves Recuperator Precooler Vessel Cooling System (VCS) Reactor Reactor Core Turbine Power conversion vessel Compressor Generator Cooling by radiation Cool down Contain Decay heat removal by passive heat transfer using the Vessel Cooling System (VCS) Ceramic-coated particle fuel Concrete containment vessel without high-pressure low-leakage performance Decay heat is removed passively from the outside of RPV by radiation without AC power. Because fuel temperature is below the limit, there is no FP release. Fuel region Lower reflector region Temperature ( o C) Upper reflector region Limit Fuel Temperature distribution at depressurization accident Max. fuel temp. < Limit temp. (1600 o C) 20

22 Safety of GTHTR300 for Blackout spent fuel storage- Temperature ( o C) Decay heat during the storage is small due to low power density Dry storage (AC power is unnecessary) 700 Cooling air Storage rack Shielding plug Spent fuel (center) Spent fuel (8 blocks/rack) Cooling air (outlet) Storage rack (surface) Time from the reactor shutdown (day) 21

23 Safety Demonstration Test using the HTTR (3/4) Reactivity (Δk/k) 2.0E E-03 R-total R-fuel R-mod. R-Xe 1.0E E E E E-03 Xenon builds up Re-criticality Reactor power peak Xenon decay -1.5E E-03 Sub-criticality Elapsed Time (h) Analytical evaluation of reactor transient during the LOFC test 22