Safety Design of HTGR by JAEA in the light of the Fukushima Daiichi accident

Transcription

1 Technical Meeting on the Safety of High Temperature Gas Cooled Reactors in the Light of the Fukushima Daiichi Accident, 8-11 April 2014, IAEA Head quarters, Vienna, Austria Safety Design of HTGR by JAEA in the light of the Fukushima Daiichi accident Hirofumi OHASHI Nuclear Hydrogen and Heat Application Research Center Japan Atomic Energy Agency (JAEA)

2 Category of safety features R&D on Safety active Passive Inherent Research Committee under AESJ ( 13-14) Development of safety requirement for commercial HTGR design NSHTR* Safety review of HTTR HTTR GTHTR300 GTHTR300C (HTGR hydrogen system) Assure the safety at accidents by inherent features Design study was initiated after F1 accident Safety requirements were developed taking into account HTGR features Reviewed and authorized by MEXT Designed by passive and inherent features Safety design philosophy for HTGR coupled with non-nuclear facility was developed * NSHTR: Naturally safe HTGR Deployment year

3 Outline 1. Event selection of Beyond Design Basis Accident in HTTR for new regulatory requirements 2. Development of new safety concept after the Fukushima Daiichi Accident 3. Summary 2

4 Background and Points for the New Regulatory Requirements Source: 3

5 New Regulatory Requirements for Research and Test Reactors issued in 18 December, 2013 Key issues Design seismic wave with significantly high acceleration External hazards BDBA and severe accident management Approach (Tentative) Re-evaluation of seismic design classification, and seismic evaluation. Evaluation of integrity against earthquake, tsunami, tornado, volcano, external fire etc. Event selection and evaluation of BDBA. 4

6 New Regulatory Requirements for BDBAs (1/3) (Preventing escalation of accidents which release large amount of radioactive materials, etc.) Article 53 The measures to prevent the escalation of accidents whose occurrence frequency is lower than that of DBAs and which have possibility of large release of radioactive materials or emission of radiation shall be considered for the research and test reactor facilities. Interpretation Article 53 (Preventing escalation of accidents which release large amount of radioactive materials, etc.) 1 Requiring evaluation and provisions for accidents whose occurrence frequency is less than that of design basis accidents and which have possibility of excess radiation exposure whose effective dose to surrounding publics of site is more than 5 msv/occurrence. 2 As regards to the postulation of the accidents, multiple failure events caused by external event such as natural hazards, etc. causing a common cause failure shall be considered. 5

7 New Regulatory Requirements for BDBAs (2/3) Interpretation 3 Example of accidents are as follows. 3-1 Accidents which cause the fuel element damage a)oxidation of core internal structure (graphite) caused by the large amount of air or water ingress than DBA, explosion of flammable gas, etc. 3-2 Accidents which cause the damage of spent fuels by loss of cooling of spent fuel storage facilities a)accidents having possibility to progress to spent fuels damage according to failure of cooling system, lack of cooling water supply, evaporation of cooling water and failure of maintaining the amount of water cooling spent fuels in spent fuel storage facilities. b)accidents having possibility to progress to spent fuel damage according to rupture of pipes of cooling system, siphon phenomenon, etc. and failure of maintaining the amount of water cooling spent fuels. c)accidents having possibility to progress to spent fuels damage according to failure of ventilating and air conditioning system in buildings and loss of cooling function as regards spent fuel storage facilities where spent fuels are cooled by air. 6

8 New Regulatory Requirements for BDBAs (3/3) 4 Measures to prevent the escalation of accidents required in Article 53 are the installation of components and planning of procedure, etc. to prevent the occurrence and escalation of accidents and to mitigate the effect of the release of radioactive materials. These are as follows for example, or measures which are equivalent or more effective to those. 4-1 Cases that the fuel element damage is postulated a)the measures to prevent the explosion of flammable gases caused by the large amount of air or water ingress into the reactor pressure vessel than DBAs by exhausting flammable gas, etc. 4-2 Cases that the spent fuels damage caused by loss of cooling function in spent fuel storage facilities is postulated a)protective provisions against failure of spent fuels, etc. by alternative facilities for pouring water (line of pouring water, fire truck, etc.), etc.. b)provisions for securing the amount of water which is enough for radiation shielding by alternative facilities for pouring water as regards spent fuel storage facilities whose radiation is shielded with water. c)provisions for maintaining subcriticality of spent fuel, etc.. d)provisions for reducing the amount of radioactive materials release into environment as possible in case of spent fuel damaged, etc.. 7

9 Methodology of Identification of BDBAs in HTTR The identification of BDBA starts with a systematic deterministic approach based on a list of AOOs and DBAs (i.e., postulating single initiating events). (1) Active safety systems to fulfill the related safety functions for theses AOOs and DBAs are identified. (2) Postulated event sequences of common cause failure or inefficiency of all redundant trains of a safety system needed to fulfill a safety function necessary to cope with an AOO or a DBA are identified by the event tree analysis. (3) Event sequences are categorized into some event sequence groups taking into account the type of loss of a safety function. (4) A bounding event sequences, which present the greatest challenges to the acceptance criteria, are selected from each event sequence group using experience feedback and engineering judgments. 8

10 AOOs in HTTR AOO AOO-1 AOO-2 AOO-3 AOO-4 AOO-5 AOO-6 AOO-7 AOO-8 AOO-9 Abnormal control rod withdrawal under subciritical condition Abnormal control rod withdrawal during rated operation Decrease in primary coolant flow rate Increase in primary coolant flow rate Decrease in heat removal by secondary cooling system Increase in heat removal by secondary cooling system Loss of off-site power Abnormality of irradiation specimens and experimental equipment Abnormality during safety demonstration tests Ref.:K. Kunitomi, et al., Safety evaluation of the HTTR, Nuclear Engineering and Design, 233, p (2004). 9

11 DBAs in HTTR DBA ACD-1 ACD-2 ACD-3 ACD-4 ACD-5 ACD-6 ACD-7 ACD-8 ACD-9 ACD-10 ACD-11 Channel blockage in fuel block Rupture of inner pipe of coaxial double pipes in primary cooling system Rupture of inner pipe of coaxial double pipes in secondary cooling system Rupture of coaxial double pipes in secondary cooling system Rupture of pipe in pressurized water cooling system Rupture of coaxial double pipes in primary cooling system (DLOCF) Rupture of heat transfer tube in pressurized water cooler Rupture of pipe in primary coolant purification system Rupture of pipe in processing facilities of radioactive gaseous waste Rupture of sweep gas pipe in irradiation test facilities Rupture of standpipe 10

12 Active safety systems to fulfill the related safety functions for AOOs and DBAs in HTTR Category of safety functions Control of reactivity Removal of heat from the reactor Confinement of radioactive material Mitigation of air ingress Mitigation of water ingress Support service systems Fast shutdown Auxiliary cooling system (ACS) Vessel cooling system (VCS) Active safety system Emergency air purification system Prevention of start-up of auxiliary cooling system Gas circulator shutoff by damping control at reactor scram Isolation valve for pressurized water system Emergency power supply Component cooling water system Compressed air supply system ACS VCS VCS Emergency air purification system 11

13 Event Tree Analysis (example) Rupture of coaxial double pipes in primary cooling system (DLOCF) Postulated initiating event DLOFC Control of reactivity Fast shutdown Mitigation of air ingress Prevention of start-up of ACS Heat removal Heat removal by VCS Confinement Emergency air purification system Event sequence (Event sequence group) DBA DLOFC + loss of Emergency air purification system (Loss of mitigation of radioactive material release) DLOFC + loss of VCS (Loss of core cooling) Success Failure DLOFC + compressed air supply system (Loss of core cooling and mitigation of radioactive material release ) DLOFC + failure of prevention of start-up of ACS (Loss of mitigation of air ingress) DLOFC + failure of fast shutdown (Loss of fast shutdown) 12

14 Event Tree Analysis (example) Rupture of heat transfer tube in pressurized water cooler (PPWCTR) Postulated initiating event PPWCTR Control of reactivity Fast shutdown Mitigation of water ingress Heat removal Confinement Gas circulator shutoff by damping control Isolation of PPWC Heat removal by ACS Emergency air purification system Event sequence (Event sequence group) DBA PPWCTR + loss of Emergency air purification system (Loss of mitigation of radioactive material release) PPWCTR + loss of ACS (Loss of core cooling) Success Failure PPWCTR + failure of isolation of water system (Loss of mitigation of water ingress) PPWCTR + failure of gas circulator shutoff by damping control (Loss of mitigation of water ingress) PPWCTR + failure of fast shutdown (Loss of fast shutdown) 13

15 Event Tree Analysis (example) Loss of off-site power Postulated initiating event Control of reactivity Heat removal Loss of off-site power Fast shutdown Heat removal by ACS Heat removal by VCS Event sequence (Event sequence group) AOO Loss of off-site power + loss of VCS (Loss of core cooling) Loss of off-site power + loss of ACS (Loss of core cooling) Success Failure Loss of off-site power + loss of emergency power supply OR loss of component cooling water system (Station black out) Loss of off-site power + failure of fast shutdown (Loss of fast shutdown) 14

16 Event Sequence Groups Postulated initiating event Loss of a safety system (Postulated multiple failure of a safety system) Event sequence group AOOs and DBAs Fast shutdown Loss of fast shutdown Auxiliary cooling system (ACS) AOOs and DBAs Vessel cooling system (VCS) Loss of core cooling Component cooling water system (ACS +VCA) DBAs Emergency air purification system Loss of mitigation of radioactive material release Air ingress accident Prevention of start-up of auxiliary cooling system at DLOFC Loss of mitigation of air ingress Water ingress accident Gas circulator shutoff by damping control at reactor scram Isolation valve for pressurized water system Loss of mitigation of water ingress Loss of off-site power Emergency power supply Component cooling water system Station blackout DBAs Compressed air supply system Loss of core cooling and mitigation of radioactive material release 15

17 Selected Event Sequences (tentative) Event sequence group Identified event sequence by event tree analysis Selected event sequence (tentative) Loss of fast shutdown AOO or DBA + failure of fast shutdown (20 events) DLOFC + failure of fast shutdown Loss of core cooling Loss of mitigation of radioactive material release Loss of mitigation of air ingress Loss of mitigation of water ingress Station blackout Loss of core cooling and mitigation of radioactive material release AOO or DBA + loss ACS (17 events) DBA + loss of VCS (3 events) AOO or DBA + Component cooling water system (20 events) AOO or DBA + loss of emergency air purification system DLOCF + failure of prevention of start-up of auxiliary cooling system Rupture of standpipe + failure of prevention of start-up of auxiliary cooling system Rupture of heat tube in pressurized water cooler + failure of gas circulator shutoff by damping control Rupture of heat tube in pressurized water cooler + failure of isolation for pressurized water system Loss of off-site power + loss of emergency power supply Loss of off-site power + loss of component cooling water system DBA + loss of VCS + loss of emergency air purification system (3 events) DLOFC + loss of VCS Rupture of pipe in primary coolant purification system + loss of emergency air purification system DLOFC + failure of prevention of start-up of ACS Rupture of heat tube in pressurized water cooler + failure gas circulator shutoff by damping control Loss of off-site power + loss of emergency power supply DLOFC + loss of VCS + loss of emergency air purification system 16

18 Outline 1. Event selection of Beyond Design Basis Accident in HTTR for new regulatory requirements 2. Development of new safety concept after the Fukushima Daiichi Accident 3. Summary 17

19 Introduction The current designed HTGRs can assure its safety without significant large release of radioactive material to the environment even in the severe conditions. That is, HTGR can achieve the regulatory requirements and safety goal under the accidents required by regulatory. In the Fukushima Daiichi accident, the multiple failure of the safety SSCs happed and large amount of radioactive materials were released to the environment. The frequency of occurrence of the multiple failures of the safety-class SSCs in the HTGR must be lower than that of other type of reactors and may be lower than that required by regulation due to its inherent characteristics and employing reliable passive safety features. After the Fukushima Daiichi accident, we feel the necessity to show the safety of HTGR in the case of multiple failure of safety-class SSCs to the future HTGR users and the nuclear experts in other fields even though its frequency of occurrence is lower than the regulatory requirements. 18

20 Introduction (continued) However, the low-frequency accident sequences have a large uncertainty. We also feel the necessity to show the more understandable reactor safety to the general public. Therefore, we have proposed the safety design concept which utilizes the HTGR inherent safety features more than the HTGR designed before the Fukushima Daiichi accident. The safety concept is the followings; The coated fuel particle is only the physical barrier to be expected the containment function for the radioactive material in the case of accidents. The coated fuel particle is protected from the challenges only by the physical phenomena in the case of accidents. * This concept dose not means no safety systems are installed. The Defence in Depth to prevent the occurrence of abnormal events and escalation of abnormal events to accidents is performed. The HTGR employing this concept is named as Naturally Safe HTGR (NSHTR). 19

21 Safety design concept of Naturally Safe HTGR (NSHTR) The integrity of physical barriers for the containment of radioactive material in the case of accidents are assured by physical phenomena. Confinement function Reactor Phenomena which degrades confinement function Causes of events Physical phenomena Fuel coating Diffusive release Sublimation Fission product Uranium Corrosion Failure Core heat up Oxidation Doppler effect Radiation, Natural convection Oxide layer formation Attain stable state Retain radionuclides within confinement RPV CV Fuel coating Flammable gas explosion Flammable gas oxidation Retain radionuclide within coated particles which are insusceptible to external events * Confinement function is only degraded by well-known phenomena * Protect confinement function only by physical phenomena * The occurrence of phenomena is unquestionable Ref: H. Ohashi et al., Concept of an Inherently-safe High Temperature Gas-cooled Reactor, American Institute of Physics Conference Proceedings, 1448, p (2012). 20

22 Physical phenomena in NSHTR - to prevent the progression of core heat up - The progression of the core heat up is prevented by inherent shutdown by the doppler effect heat removal from the outside of RPV to soil by conduction in the RPV, radiation and natural convection of air in the reactor cavity HTTR (Active) Heat GTHTR300 (Passive) Heat NSHTR (Inherent) VCS M Natural convection Natural convection Natural convection conduction Radiation Forced circulation conduction Radiation Natural convection conduction Radiation Heat VCS: Vessel Cooling System 21

23 Physical phenomena in NSHTR - to prevent the progression of coating layer oxidation - Initial O 2 Pressure (Pa) The progression of SiC oxidation is prevented by SiO 2 layer formed on SiC surface Temperature (K) Passive oxidation SiC + C SiC Active oxidation / T (10-4 /K) O 2 OPyC SiO 2 SiC IPyC Cross section CFP Passive oxidation SiC(s) + 3/2O 2 (g) SiO 2 (s) + CO(g) Active oxidation SiC(s) + O 2 (g) SiO(g) + CO(g) Ref: International Atomic Energy Agency, Response of Fuel, Fuel Elements and Gas Cooled Reactor Cores under Accidental Air or Water Ingress Conditions, IAEA-TECDOC-784, IAEA, Vienna (1995). 22

24 Physical phenomena in NSHTR - to prevent the progression of the flammable gas explosion - CO concentration is maintained below the explosion limit by CO oxidation reaction. CO oxidation CO + 1/2O 2 CO 2 O C O Graphite Outlet Pipe rapture Air ingress (Natural convection) Graphite oxidation C +1/2O 2 CO C + O 2 CO 2 C Inlet Coolant channel % Lower limit of explosion CO concentration 23

25 Feasibility of NSHTR concept - Core heat up, CO explosion (1/2) Core temperature ( o C) CO concentration (mol%) x CO production Decay heat CO oxidation Conduction Natural convection Conduction CO explosion limit:12.5% 8 6 Equivalent diameter :15mm 4 At core outlet Graphite temperature [ o C] Core temp. limit:1600 o C RPV Oxidation heat x Air ingress Radiation Surrounding structures Core radius:2.76 m : 5.3 MW/m 3 (600MW), Decay heat : 4.2 MW/m 3 (480MW), Decay heat + Graphite/CO oxidation heat Elapsed time (h) 24

26 Feasibility of NSHTR concept - Core heat up, CO explosion (2/2) Power density (MW/m 3 ) Equivalent diameter (mm) Core heatup Conduction Model Natural convection Thermal radiation Core RPV Surroundings Design capable Energy balanced equation in lumped parameter system Cylindrical composite Decay heat ANS 5.1 including Actinide decay Heat sink temp.: 30 o C Core temp. limit: 1600 o C Heat sink Design criteria for core heat removal : Decay heat : Decay heat + Graphite/CO oxidation heat Core radius (m) Graphite oxidation, CO explosion Model O C O C Air ingress CO oxidation CO + 1/2O 2 CO 2 Graphite oxidation C + 1/2O 2 CO C + O 2 CO 2 Diffusion in the pore and mass transfer Natural convection Graphite block Mass, energy and momentum conservations for steady-state, onedimensional and multicomponent system Circular tube Graphite and CO oxidation heat Graphite temp.: 500~ 1600 o C Design criteria for oxidation heat and CO explosion Graphite/CO oxidation heat can be removed Design capable Below CO explosion limit Total cross-sectional area of flow channel (m 2 ) 25

27 R&D items for NSHTR Phenomena which degrades confinement function Diffusion Sublimation Confinement function of CFP Heat up Cause of events Increase of heat generation Decrease of cooling Reactivity insertion Flow rate increase R&D on NSHTR Design study Development of safety requirements HTTR test (reactor physics) Corrosion Oxidation Air ingress DLOFC HTTR test (Fission product) Failure Explosion Air ingress Out-of-pile test Fuel oxidation Air ingress 26

28 3. Summary The evaluation of BDBAs in HTTR is underway for the safety review by Nuclear Regulation Authority according to new regulatory requirements for research reactors. JAEA has started the study on Naturally Safe HTGR, which ensure no large FP release by inherent features, after the Fukushima Daiichi accident. 27