T. Nakatsuka , Y. Oka , Y. Ishiwatari , S.Ikejiri , K. Okumura , S. Nagasaki , K. Tezuka , H. Mori , K. Ezato , N. Akasaka , Y. Nakazono , K.

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1 Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors 5 to 8 July 2010, Pisa, Italy 1 T. Nakatsuka 2, Y. Oka 1,*, Y. Ishiwatari 1, S.Ikejiri 1,* *, K. Okumura 2, S. Nagasaki 1, K. Tezuka 3, H. Mori 4, K. Ezato 2, N. Akasaka 2, Y. Nakazono 1, K.Sasaki 1, T. Terai 1, Y. Muroya 1 and M. Yamakawa 1 Present study is the result of Research and Development of the Super Fast Reactor entrusted to The University of Tokyo by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).

2 Japanese SCWR Program 2 Two R&D projects on the pressure vessel type SCWR with fast/thermal options 1. Fast option (Dec Mar. 2010) " Research and development of the Super Fast Reactor" Entrusted to the University of Tokyo Funded by MEXT 2. Thermal option (Aug Mar. 2011) Development of SCWR in GIF Collaboration (Phase I), Toshiba and The Institute of Applied Energy Funded by METI

3 Introduction of preceding studies 3

4 Objectives To study on the conceptual design of Super FR which have high thermal efficiency and obtain the compactness and simplicity of the plant due to the high specific enthalpy of supercritical pressure light water. To develop a data base for heat transfer coefficients of supercritical water and material properties by carrying out the experiments. Control Rods Supercritical water Reactor Pomp Super FR Turbine Generator Condenser Heat sink 4 Electric Power Source:

5 Research subjects 5

6 Subtheme Conceptual design of Plant To make clear Concept and Features of Super FR from the design under evaluating the interaction between Core/Fuel/High temperature Structural designs and Plant Control/Safety analyses. 6

7 Subtheme Conceptual design of Plant 7

8 3-D coupled core neutronic/thermal-hydraulic calculation MLHGR[kW/m] MLHGR MCST MCST[ o C] Cycle burnup exposure[gwd/t]

9 Design of Improved Core 9 Improved core is designed from resolving the issues obtained by safety and fuel analyses as shown in the following method. 3-D 3-D Nuclear/Thermal coupled calculation Safety Analysis Fuel Analysis Improvement of Core specifications smaller fuel rod diameter Improved Core

10 Fuel rod and core design 10 The core of the Super Fast Reactor consists of seed and blanket fuel assemblies. MOX and stainless steel cladding are used for seed fuel rods. Seed fuel assembly Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

11 Fuel rod and core design 11 Depleted UO 2 and stainless steel cladding are used for blanket fuel rods. The fuel rod region is surrounded by a solid moderator (ZrH layer). Fast neutrons coming from the seed fuel slow down in the ZrH layer and are absorbed by the blanket fuel without causing fast fissions. It enables the Super Fast Reactor to have a negative void reactivity without adopting flat core shape or additional devices. Fuel rod Stainless steel ZrH layer Blanket fuel assembly Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

12 Fuel pin and improved core 12 Improved core can be designed with not only achieving smaller fuel rod diameter and higher power density but also satisfying safety criteria. Reference Core Improved Core Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

13 Fuel pin and improved core 13

14 Summary of core design 14

15 High Temperature Structural design 15 Shape and dimensions of pressure vessel, reactor structure and reactor internal structures have been determined corresponding the improved core design. Blanket assembly Seed assembly (Fresh) Seed assembly (2nd cycle) Seed assembly (3rd cycle) Seed Blanket assembly assembly Reflector assembly H17 H21 Reference design Improved design Cross-section view of Reactor vessel

16 High Temperature Structural design 16 Confirmed the structural integrity can be ensured in the new structural design from the analyses on steady and transient thermal stress of (H21 steam outlet nozzle and upper tie plate in referring to the results of the safety analysis. Top dome Steam outlet nozzle CR guide Upper tie plate RPV inlet (Inlet shifted 90 degrees) Core outlet Reactor upper structure

17 Plant system and safety system 17 Control system was designed by using the plant transient code. Safety system was designed by considering Super FR features ; once through direct cycle high power density small reactivity feedback of coolant density

18 Plant system and safety system 18 Stand-by liquid control system (SLC) Containment Vessel Reactor Pressure Vessel Core spray system MSIV Control rod drive Safety relief valve (SRV) Automatic depressurization Turbine control valve system (ADS) Core Main spray steam isolatio system n valve Turbine bypass valve Turbin e Condenser Low pressure condensa te pump Auxiliary feed water system (AFS) Low pressure coolant injection (LPCI) Condensate water storage tank Driving Turbin e High pressure feed water Heaters Booster pump Low pressur e feed water Heaters Reactor coolant pump (Main feed water pump) Deaerator High pressure condensa te pump Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

19 Maximum cladding temperature Analysis results of maximum cladding temperature for each event Safety Criteria Safety Criteria Accident Partial Events Accident Abnormal transient Events Events Event Number ATWS: Anticipated Transient Without Scrum Analysis Loss of feed water power Partial Loss of coolant Event Loss of offsite power Abnormal Loss of feed water Loss of load (OFF Turbine bypass transient Abnormal power False start of Auxiliary feed Anticipated valve) withdrawal of Control rod (Stat transient Abnormal water Partial Loss of coolant up) Anticipated withdrawal of Control rod Abnormal transient (Operation) transient Abnormal Loss of offsite power False start of main feed water Loss of load (ON Turbine bypass False systemclose of Main steam isolation transient Abnormal Loss valve) of load (OFF Turbine bypass Accident valve Loss of coolant transient Abnormal Anticipated valve) withdrawal of Control rod (Stat Event Accident Stack of main pump rotating transient Abnormal Anticipated up) withdrawal of Control rod Event Accident Pull axis out of control rod on full power transient Abnormal (Operation) False start of main feed water Accident Event operation Pull out of control rod on high temperature Abnormal transient Failure systemof core pressure control Accident Event Event transient Abnormal system Rupture waiting of main pipe Loss of coolant False close of Main steam isolation Accident (LOCA) Event transient valve (LOCA) Rupture of main pipe Loss of coolant 19 (LOCA)

20 Comparison of plant size 20

21 Subtheme (2) Thermal Hydraulics 21 HCFC22 Loop at the Kyushu Univ. Supercritical pressure water loop (JAEA) Heater rod With grid spacer

22 Experiments with Supercritical Pressure Freon 22 The test section geometry started from a smooth circular tube and shifted to a 7 rod bundle. To evaluate the effect of the grid spacers on the heat transfer, several types of grid spacers had been tested. Wall temperature Tw C Heat transfer coefficient α kw/(m 2 K) HCFC22 Upward flow T b - h pc T pc P =5.5MPa G = 1000 kg/(m 2 s) q =40kW/m Bulk fluid enthalpy h b kj/kg h pc Dittus-Boelter

23 Experiments with Freon 23 The critical heat flux (CHF) and the post boilingtransition (BT) heat transfer were also measured since those are important, especially just below the critical pressure. If sliding pressure startup scheme is adopted, the Super Fast Reactor is operated at subcriticalpressure during startup. The pressure decreases from supercritical to subcritical at pressure decreasing events including LOCA.

24 Discharge & Condensation Test 24 The measurement of pressure oscillation and the observation of condensation behavior of a vapor discharge into sub cooled liquid pool has been carried out to obtain basic database for the evaluation of safety of the LOCA in the Super Fast Reactor. A surrogate fluid (HCFC 123) is used. The vapor at high temperature is discharged into the sub cooled liquid pool through a submerged single pipe vertically fixed. The oscillatory condensation is observed. The condensation oscillation produces pressure oscillation in the liquid pool. The condensing interface area becomes small as the increase of the degree of sub cooling. The pressure frequency has a period of millisecond order and the frequency and amplitude of the pressure oscillation increase with increasing the degree of sub cooling and mass flux of the vapor, like the results of some conventional water vapor injection tests Pressure amplitude P rms [kpa] Vapor temperature 100 [ ] Pipe dia 10 [mm] Back pressure 370 [kpa] Mass flux [kg/m 2 s] Liquid subcooling [K] Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

25 Experiments with Supercritical Pressure Water 25 A single heater rod and a 7 rod bundle with grid spacers, which are the same type as those of the loop of the Kyushu University for HCFC22, were tested. Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

26 Development of CFD code based on Two Fluid model 26 The existing 3 D CFD code ACE 3D, which has been developed by JAEA based on the twofluid model, was enhanced to handle the supercritical pressure fluids in the fuel bundle geometry. Prediction accuracy of the code has been improved by comparing with experimental results with Freon and water. Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

27 Sub Theme(3) Development of Core Material 27 Trial Production of Materials -For improvement of high temperature strength and corrosion resistance Evaluation of Experimental Material -Tensile test -Creep test -Corrosion test -SCC susceptibility testing (SSRT Test) -Ion irradiation test -Irradiation test under complex environment Evaluation of neutron irradiated material -Corrosion test -SCC susceptibility testing (SSRT Test)

28 Fuel cladding material 28 Ti added advanced 15Cr 20Ni steel and Zradded advanced 15Cr 20Ni steel based on PNC1520 that is an austenitic stainless steel for FBR claddings having excellent creep strength at high temperature are produced experimentally. Strength test and the corrosion test on these materials are executed.

29 Fuel cladding material 29 The Zr added advanced 15Cr 20Ni steel has excellent characteristic on corrosion resistance and strength as a result of creep test in air and corrosion test in SCW (600 O C, 25MPa) and high temperature high pressure water (290 O C, 8MPa). Creep strength of candidate material Weight loss of candidate material Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

30 Development of highly durable thermal shielding material 30 Thermal insulations will be necessary for the coolant guide tubes attached to the blanket fuel assemblies of the Super Fast Reactor due to large temperature difference in the reactor vessel. The requirements for the thermal insulator are: a low heat conductivity, a low neutron absorption, a good thermal shock resistance and dimensional stabilities.

31 Development of highly durable thermal shielding material 31 High performance thermal shielding material (Yittria stabilized zirconia) has been developed by optimization of the pore former and the sintering temperature. sintering density: 25% of the dense zirconia thermal conductivity: 1/20 of the dense zirconia Thermal conductivity of the sintered porous YSZ. (sintering density = 25%) 3YSZ: 3mol% Y 2 O 3 97mol% ZrO2 8YSZ: 8mol% Y 2 O 3 92mol% ZrO2 Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

32 Development of highly durable thermal shielding material 32 Yittria stabilized zirconia Performing a comprehensive evaluation of the relation between a thermal conductivity and mechanical strength. Compressive strength as a function of sintering temperature. 3YSZ: 3mol% Y 2 O 3 97mol% ZrO2 8YSZ: 8mol% Y 2 O 3 92mol% ZrO2 Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

33 Elution properties of stainless materials in SCW 33 It is very important to understand the elution characteristic of stainless materials in supercriticalpressure water in order to manage turbine contamination. A new approach was developed to evaluate the elution characteristic of stainless materials in supercriticalpressure water. Radioactivated specimen (SUS304) with a known radioactivity is used as sample set at an autoclave vessel in a supercritical water loop system. By measuring gamma ray emitted from the eluted material, the elution can be quantified with high sensitivity. Meas. for WQ Purifier for WQ Meas. for WQ Purifier for WQ Pump Drain Ar, H 2, O 2 HP Pump Pump Tank for Chemical s Water Tank γ-ray detector (NaI) Heater Pre-heater Cooler Ion exchange resin Radioactivated specimen (SUS304) Autoclave T: C P: 25 MPa Source: Z. HAN and Y. MUROYA, A New Method to Evaluate the Elution and Corrosion Properties of Stainless Materials in Subcritical and Supercritical Water, Symposium on Water Chemistry and Corrosion of Nuclear Power Plants in Asia, (2009) RT

34 Elution properties of stainless materials in SCW 34 The elution properties of stainless materials in SCW was investigated by usage of a SCW loop system. The elution is extremely suppressed at high temperatures, especially in supercritical state. Elution efficiency [g/m 2 ] No data DH:40ppb 300C 400C 450C 550C Deaerated DO:200ppb DO:400ppb Conditions No data No data No data Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

35 Elution properties of stainless materials in SCW 35 Water properties change dominates the elution properties at rapid temperature shift condition. Elution effiency [g/m 2 ] Initial oxide layer efficiently suppresses the elution o C 300 o C Time [hr] Source: T. Nakatsuka et al., Proc. Technical Meeting on Heat transfer, thermal-hydraulics and system design for supercritical pressure water cooled reactors, TM-18 (2010)

36 36 Business articles have been performed in FY and then the initial goals are achieved as initial plan. In 1st subtheme Plant concept design, concept and features of once through direct cycle supercritical pressure light water cooled Fast Reactor are quantitatively cleared from overall design procedure of fuel, core, high temperature structural design and safety analysis. In 2nd subtheme R&D of core thermal hydraulics, through performing 7 pin bundle test etc., basic characteristics of thermal hydraulics is cleared and then it s basic data base is made.

37 In 3rd subtheme Development of high strength cladding materials etc., improved PNC 1520 materials were tested under the mixed conditions of irradiation and corrosion and then the characteristics of candidate materials were cleared. It is clarified that Zr added advanced 15Cr 20Ni steel has excellent characteristics. Yttria stabilized zirconia having excellent insulation characteristics has been developed and it s characteristics about strength and corrosion in supercritical pressure light water etc. are collected. Combined effects of water temperature and dissolved oxygen about elution characteristics into water is evaluated for core materials and that outlines are clarified. These results will be used at international activities as Generation IV International Forum GIF for not only using as infrastructure of future R&D but also leading worldwide SCWR R&D. 37

38 38 Monograph Super LWR and Super Fast Reactor (Y. Oka et al.) will be published in July 2010 from Springer. Present study is the results of Research and Development of the Super Fast Reactor entrusted to the University of Tokyo by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).

39 39 Future work ( )