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CONCEPTUAL DESIGN OF THE HTTR-IS NUCLEAR HYDROGEN PRODUCTION SYSTEM On the other hand, the IS process in the HTTR-IS system should be built outside and set at an appropriate distance from the reactor building in order to prevent flammable and toxic gas inflow to the building from the reactor safety point of view (Sakaba, 2008). Hence, the secondary cooling system will be modified to penetrate the CV and reactor building in order to provide heat to the IS process. The IHXTR in the HTTR-IS system would result in radionuclide transportation to the outside of the CV. In order to reduce radionuclide transportation from the primary to the secondary cooling system, containment isolation valves should be installed. These valves will be automatically closed by the signal detecting the heat transfer tube rupture of IHX. Meanwhile, the current engineered safety features actuating system is not designed to detect the rupture since the scenario did not impact on the reactor safety of the original HTTR. Therefore, a detection method of heat transfer tube rupture of IHX should be established. Detection method of IHX tube rupture The strategy of detection method establishment is to apply the simple and reliable method since it would comprise the engineered safety featured actuation system. During the IHXTR, helium flows through the rupture point from the secondary to the primary cooling system until the pressure of both balances. In the meanwhile, pressure control systems try to maintain the set pressures of cooling systems. The following parameters would vary sensitively: • differential pressure between primary and secondary cooling system; • primary helium gas recovery flow rate due to primary helium pressure control system; • secondary helium gas supply flow rate due to primary-secondary differential pressure control system. Monitoring differential pressure between the primary and secondary cooling systems would be one of the candidates for detecting the heat transfer tube rupture. Meanwhile, the pressure varies even at normal operation and therefore there is a concern of malfunction of CV isolation valves. In case of primary helium gas recovery flow rate, primary helium pressure control system only covers the rated operation since the system activates when the primary coolant pressure reaches about 3.95 MPa. On the other hand, the primary-secondary differential pressure control system covers start-up, shutdown and rated operations. The control system keeps supplying helium gas to the secondary cooling system during the scenario. Thus, monitoring the entire secondary helium gas supply would be an effective way to detect the tube rupture. Numerical evaluation of the detection methods The availability of the proposed detection method was evaluated. Furthermore, system behaviour is assessed during the scenario. A system analysis code developed for VHTR systems is used for these calculations. Outline of the system analysis code JAEA conducted an improvement of the RELAP5 MOD3 code (US NRC, 1995), the system analysis code originally developed for LWR systems, to extend its applicability to VHTR systems (Takamatsu, 2004). Also, a chemistry model for the IS process was incorporated into the code to evaluate the dynamic characteristics of process heat exchangers in the IS process (Sato, 2007). The code covers reactor power behaviour, thermal-hydraulics of helium gases, thermal-hydraulics of the two-phase steam-water mixture, chemical reactions in the process heat exchangers and control system characteristics. Field equations consist of mass continuity, momentum conservation and energy conservation with a two-fluid model and reactor power is calculated by point reactor kinetics equations. The code was validated by the experimental data obtained by the HTTR operations and mock-up test facility (Takamatsu, 2004; Ohashi, 2006). Figure 3 shows the nodalisation of the HTTR-IS system model. The reactor consists of the internal flow path (P2), permanent reflector blocks (HS25), upper plenum (B4), reactor pressure vessel (RPV) (HS30), vessel cooling system, reactor core bypass flow (P10), lower plenum (B12) and reactor core. The 390 NUCLEAR PRODUCTION OF HYDROGEN – © OECD/NEA 2010
CONCEPTUAL DESIGN OF THE HTTR-IS NUCLEAR HYDROGEN PRODUCTION SYSTEM Figure 3: Nodalisation of the HTTR-IS system model TV 190 9 8 T J1 TV 184 TJ185 0 P 5 AHX 92 52 P: Pipe SV: Single volume B: Branch TV: Time dependent volume TJ: Time dependent junction V: Valve Pump, Circulator Heat structure P94 P54 P48 TJ187 TV 186 1 0 P B4 0 P 1 8 P B12 Reactor P56 2 P B1 V C S 8 P 2 1 6 P B14 P24 1 8 P 6 P 8 6 2 P Containment isolation valves P88 P166 P90 V91 P164 P64 P60 V77 6 8 1 8 1 B P 1 V182 TV 183 P22 IHX 20 P84 P46 P44 42 P38 82 36 P76 74 P80 P40 P34 P32 5 P102 P170 B140 1 2 B P171 P172 P702 P174 P704 P96 P706 P98 P708 P160 TV 178 V179 B180 B72 PPWC TV 150 P152 P68 B70 TV 385 Helium cooler Process heat exchangers TV 142 3 7 0 S V 3 8 0 S V 3 9 0 S V TV 154 P162 Air 5 4 2 S V P360 SV355 B280 SV270 SV260 B250 B244 SV242 SV240 SV232 SV232 SV222 SV220 P66 B210 B114 3 4 5 B Air cooler TV 290 TV 200 3 0 S V Steam generator 0 P 3 6 1 P P30 B104 P100 B132 B126 B130 124 128 B120 reactor core is classified into two sections: a hot channel with a flow channel (P6), fuel (HS10) and graphite block (HS11), and average channels with a flow channel (P8), fuel (HS20) and graphite block (HS21). The flow rate of the reactor core bypass flow is adjusted by a pressure loss using valve model (V9). Radiation heat transfer in radius direction are considered and heat produced by the fuel transfers to RPV. The vessel cooling system is modelled setting a RPV surface temperature in this model. PPWC consists of helium flow (B32), pressurised water flow (P102) and heat transfer tube (HS50). The IHX consists of primary side (P18), secondary side (P62) and heat transfer tube (HS40). The steam generator is a thermo-siphon type and classified into: an upper plenum (SV260, 270 and B250), heat exchange sections (SV220, 222, 230, 232, 240 and 242), a lower plenum (B210), a downcomer (P300), helium flow (P66) and heat transfer tubes (HS220, 230, 240). A mechanical-draft type air cooler is modelled considering the elevation from the steam generator and consists of a water-steam mixture flow (P360), air flow (TV401-425) and heat transfer tubes (HS360). The air flow rate is set as a boundary condition. Primary gas circulators (Pump 20, 36, 42 and 82), a secondary gas circulator (Pump 74) and pressurised water pumps (Pumps 124 and 128) are modelled and their rotational speed is controlled by primary, secondary and pressurised water flow rate control system, respectively. The primary and secondary helium storage and supply systems are modelled by storages tanks (TV178, 183, 184, 186 and 190), supply and recovery valves (V179 and 182, TJ185, 187 and 189), and actuated by pressure control systems. Containment isolation valves (V77 and 91) are modelled as motor valves and actuate by a CV isolation signal produced by a trip system. In the following calculation, a deterministic approach would be applied in order to evaluate reactor safety conservatively. Steady-state conditions were initialised based on values used in the HTTR safety review (Saito, 1994). NUCLEAR PRODUCTION OF HYDROGEN – © OECD/NEA 2010 391
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Nuclear Science 2010 Nuclear Produc
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ORGANISATION FOR ECONOMIC CO-OPERAT
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TABLE OF CONTENTS Table of contents
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TABLE OF CONTENTS Y. Gong, E. Chalk
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EXECUTIVE SUMMARY Executive summary
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EXECUTIVE SUMMARY steam turbine, in
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EXECUTIVE SUMMARY sulphur depositio
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EXECUTIVE SUMMARY affords saving 35
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EXECUTIVE SUMMARY safety assessment
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EXECUTIVE SUMMARY The question was
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CHANGING THE WORLD WITH HYDROGEN AN
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NUCLEAR HYDROGEN PRODUCTION PROGRAM
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NUCLEAR HYDROGEN PRODUCTION PROGRAM
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FRENCH RESEARCH STRATEGY TO USE NUC
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PRESENT STATUS OF HTGR AND HYDROGEN
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STATUS OF THE KOREAN NUCLEAR HYDROG
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THE CONCEPT OF NUCLEAR HYDROGEN PRO
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CANADIAN NUCLEAR HYDROGEN R&D PROGR
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APPLICATION OF NUCLEAR-PRODUCED HYD
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STATUS OF THE INL HIGH-TEMPERATURE
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HIGH-TEMPERATURE STEAM ELECTROLYSIS
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MATERIALS DEVELOPMENT FOR SOEC Mate
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A METALLIC SEAL FOR HIGH-TEMPERATUR
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DEGRADATION MECHANISMS IN SOLID OXI
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CAUSES OF DEGRADATION IN A SOLID OX
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70 μm CAUSES OF DEGRADATION IN A S
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NUCLEAR HYDROGEN USING HIGH TEMPERA
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SESSION III: THERMOCHEMICAL SULPHUR
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CEA ASSESSMENT OF THE SULPHUR-IODIN
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STATUS OF THE INERI SULPHUR-IODINE
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INFLUENCE OF HTR CORE INLET AND OUT
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EXPERIMENTAL STUDY OF THE VAPOUR-LI
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DEVELOPMENT OF HI DECOMPOSITION PRO
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SOUTH AFRICA’S NUCLEAR HYDROGEN P
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INTEGRATED LABORATORY SCALE DEMONST
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DEVELOPMENT STATUS OF THE HYBRID SU
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RECENT CANADIAN ADVANCES IN THE THE
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AN OVERVIEW OF R&D ACTIVITIES FOR T
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STUDY OF THE HYDROLYSIS REACTION OF
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DEVELOPMENT OF CuCl-HCl ELECTROLYSI
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EXERGY ANALYSIS OF THE Cu-Cl CYCLE
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CaBr 2 HYDROLYSIS FOR HBr PRODUCTIO
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SESSION V: ECONOMICS AND MARKET ANA
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DEVELOPMENT OF THE HYDROGEN ECONOMI
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NUCLEAR H 2 PRODUCTION - A UTILITY
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ALKALINE AND HIGH-TEMPERATURE ELECT
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THE PRODUCT OF HYDROGEN BY NUCLEAR
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SUSTAINABLE ELECTRICITY SUPPLY IN T
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PROHYTEC, THE FRENCH INDUSTRIAL PLA
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NHI ECONOMIC ANALYSIS OF CANDIDATE
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Page 342 and 343: NHI ECONOMIC ANALYSIS OF CANDIDATE
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Page 346 and 347: MARKET VIABILITY OF NUCLEAR HYDROGE
Page 348 and 349: POSSIBILITY OF ACTIVE CARBON RECYCL
Page 357 and 358: NUCLEAR SAFETY AND REGULATORY CONSI
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Page 380 and 381: PROPOSED CHEMICAL PLANT INITIATED A
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Page 399 and 400: USE OF PSA FOR DESIGN OF EMERGENCY
Page 409 and 410: HEAT PUMP CYCLE BY HYDROGEN-ABSORBI
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Page 435 and 436: ALTERNATE VHTR/HTE INTERFACE FOR MI
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ALTERNATE VHTR/HTE INTERFACE FOR MI
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ALTERNATE VHTR/HTE INTERFACE FOR MI
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POSTER SESSION CONTRIBUTIONS Poster
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ACTIVITIES OF NUCLEAR RESEARCH INST
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ACTIVITIES OF NUCLEAR RESEARCH INST
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A URANIUM THERMOCHEMICAL CYCLE FOR
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A URANIUM THERMOCHEMICAL CYCLE FOR
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MEETING ORGANISATION Annex 1: Meeti
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LIST OF PARTICIPANTS REYTIER, Magal
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LIST OF PARTICIPANTS BROWN, Nichola
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LIST OF PARTICIPANTS SINK, Carl US
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