Re evaluation of Maximum Fuel Temperature

Transcription

1 IAEA Technical Meeting on on Re evaluation of Maximum Operating Temperatures and Accident Conditions for High Temperature Reactor Fuel and Structural Materials, July 2012, Vienna, Austria Re evaluation of Maximum Fuel Temperature of the HTTR at Normal Operation Hirofumi OHASHI Nuclear Hydrogen and Heat Application Research Center Japan Atomic Energy Agency (JAEA)

2 Outline 1. HTTR overview 2. Evaluation of HTTR fuel temperature at design stage Evaluation method Evaluation results 3. Re evaluation of fuel temperature 1 st modification using the operation data of the riseto power test up to 850 o C 2 nd modification using new analysis model 3 rd modification using the 950 C operation data 4. Related future tests using HTTR 5. Summary 1

3 High Temperature Engineering Test Reactor (HTTR) Major specification HTTR Fuel Rods Intermediate heat exchanger (IHX) Containment vessel Graphite Block Reactor pressure vessel Hot gas duct Fuel Uranium enrichment Core Fuel assembly Moderator Primary coolant Thermal power Inlet temperature Outlet temperature Primary coolant pressure Primary coolant flow rate Low enriched UO2 3~10wt% (avg. 6wt%) Prismatic Pin in block Graphite Helium 30 MW 395 C 850oC / 950 C (Max.) 4 MPa 12.4 / 10.2 kg/s First criticality : 1998 Full power operation (850oC/30 MWt): oC operation at full power: days continuous operation at high outlet temperature (950oC/30MWt) :

4 Structure of Fuel Assembly Fuel kernel,600μm High density PyC SiC Low density PyC 39mm 8mm 920μm Coated fuel particle Plug Fuel compact Graphite sleeve Dowel pin Fuel handing hole 580mm 26mm 34mm Dowel socket 360mm Fuel compact Fuel rod Fuel assembly 3

5 Reactor Core Structure Control rod standpipe Reactor pressure vessel Permanent reflector Replaceable reflector Fuel assembly Core restraint mechanism Core support plate Primary helium gas tube Reactor pressure vessel Core restraint mechanism Permanent reflector Replaceable reflector Control rod guide block Fuel assembly 4

6 Evaluation Method of HTTR Fuel Temperature (1) Nuclear design code Power density and neutron fluence distributions Fuel, control rod, core component, core internal structure design data (2) FLOWNET In vessel thermal and hydraulic analysis code Coolant flow rate and coolant temperature distribution (3) TEMDIM Fuel temperature analysis code Fuel temperature 5

7 In vessel Thermal and Hydraulic Analysis Code FLOWNET :Gap between each block : Control rod column flow path : Fuel channel Top shielding Replaceable reflector Fuel assembly Replaceabl e reflector Hot plenum One dimensional model using nodes and branches The flow channels are represented by node, and the nodes are connected by branch. The heat transfer between the branches are taking into account. Flow paths: the main coolant flow, the bypass flow in the inter column gaps, the leakage flow through the permanent reflectors and the cross flow in the horizontal interface gaps of the hexagonal graphite blocks Ref: S. Saito et al., Design of High Temperature Engineering Test Reactor (HTTR), JAERI 1332 (1994). 6

8 Fuel Temperature Analysis Code TEMDIM Two dimensional cylindrical model based on the power distribution including local power peaking, coolant flow distribution including redistribution in the fuel column and hot spot factors Fundamental equation T gin Coolant T F i FUEL T T Gas in F i Ti i 1 m(i) 2 fs i,j 1 fri,k j 1 k 1 n(i) FUEL : Fuel temperature ( ) 5 N T 1 N T 2 N N N T 3 T 4 T 5 N T A A cross section Gas T in : Coolant inlet temperature ( ) T i F i : Temperature rising ( ) : Hot spot factor ( ) A A Estimated point fs i,j fr i,k : Random factor (e.g., manufacturing tolerances, uncertainties on physical properties) ( ) : Systematic factor (e.g. total reactor power, coolant flow rate, inlet coolant temperature )( ) i= 1 : Coolant temperature rising 2 : Film temperature rising 3 : Temperature rising in graphite sleeve 4 : Fuel compact graphite sleeve gap temperature rising 5 : Temperature rising in fuel compact Graphite block Fuel rod Fuel compact Gap Graphite sleeve Annular flow path Coolant 7

9 Evaluation Result of Fuel Temperature at Design Stage Top plenum 98.9 The effective flow rate for the fuel cooling is about 88% of total flow. Remains are the leakage flow through the permanent reflectors, flow in the control rod cooling channel, and the bypass flow in the inner column gap. Upper shield Replaceable reflector Fuel block Replaceable reflector Control rod cooling channel Fuel cooling channel Inner column gap Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). Plenum block Hot plenum (Unit:%) 100 Outlet Inlet 100 8

10 Evaluation Result of Fuel Temperature at Design Stage Nominal temp. Maximum temp Horizontal temp. distribution Core Center Vertical temp. distribution for the fuel where the maximum fuel temp. appeared. Vertical position at a fuel column (Top) (Bottom) Temperature ( o C) Graphite block Coolant Sleeve outer surface Compact inner surface (nominal) (maximum) Temperature limits of HTTR fuel Anticipated operation occurrence: 1600 C Normal Operation : 1495 C Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). 9

11 Re evaluation of HTTR Fuel Temperature (1) 1st modification using the operation data of the rise topower test up to 850 C Validation of power and helium flow distributions Revision of operating conditions (e.g., core inlet coolant temperature) Revision of hot spot factors (2) 2nd modification using new analysis model Detailed mesh model (3) 3rd modification using the 950 C operation data in 2004 Revision of core inlet coolant temperature and core coolant flow rate 10

12 Validation of Evaluation Results using HTTR data Estimated power and helium flow distributions were validated using the operation data of the rise to power test up to 850 o C. Power distribution Flow distribution F5 F4 F3 F2 F4 F2 F1 F5 F1 F3 Center The gross gamma ray from the fuel assemblies was measured by GM counter T/C for core inlet coolant temp. Center region Outside region (6 points) T/C for hot plenum coolant temp. Power density (W/cc) Measured Calculated F1 F2 F3,F4 F5 Column number Coolant temperature rising ( o C) Measured Calculated Center region Outside region Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). 11

13 1 st Modification: Modification of Calculation Conditions Boundary conditions concerning to the operating conditions, and hot spot factors (i.e., systematic factors) were revised using the operation data up to 850 o C operation. Design Revised Reasons Operation day 440 EFPD 160 EFPD Operating condition at 1 st 950 o C operation Core inlet coolant temperature 415 o C 409 o C Operation data Core coolant flow rate 10.2 kg/s 10.1 kg/s Re evaluation of heat and mass balance using operation data Control rod position 2610 mm 2900 mm Operation data Systematic factors Factor for thermal power Coolant temp. rise 2.5% 0% Factor for axial power distribution Factor for flow distribution Others 2.5% 2.0% 4.0% 0% No effect of the thermal power error on the coolant temp. rise Calibration result of thermal power Measurement results of power distribution Coolant temp. rise 4.0% 2.0% Re evaluation of flow distribution using operation Film temp. rise 3.2% 1.6% data Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). 12

14 1 st Modification: Re evaluation Result of Fuel Temperature Top plenum Design stage 98.9 Flow distribution 1 st modification Top plenum 99.6 Upper shield Replaceable reflector Fuel block Replaceable reflector Control rod cooling channel Fuel cooling channel Inner column gap Upper shield Replaceable reflector Fuel block Replaceable reflector Control rod cooling channel Fuel cooling channel Inner column gap Plenum block 99.3 Hot plenum Plenum block 99.2 Hot plenum (Unit : %) Outlet Inlet (Unit : %) Outlet Inlet Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). 13

15 1 st Modification: Re evaluation Result of Fuel Temperature Design stage Fuel temperature 1 st modification Nominal temp. Maximum temp. Nominal temp. Maximum temp Core Center Core Center Estimated maximum fuel temperature was decreased from 1492 o C at design stage to 1463 o C by the re evaluation using the operation data. Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). 14

16 1 st Modification: Re evaluation Result of Fuel Temperature Temperature distribution Design stage 1 st modification (Top) 1 Temperature ( o C) (Top) 1 Temperature ( o C) Vertical position at a fuel column Graphite block Coolant Sleeve outer surface Vertical position at a fuel column Graphite block Coolant Sleeve outer surface (Bottom) Compact inner surface (nominal) (maximum) (Bottom) Compact inner surface (nominal) (maximum) Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). 15

17 2 nd Modification: Modification of Analysis Model Old model Each fuel block is divided into 6 triangularmeshes for the models of nuclear design and the fuel temperature analysis. One fuel rod is represented by the triangularmesh of the fuel temperature analysis model. The rector power calculated by the nuclear design is allocated for each mesh of the fuel temperature analysis model using the peaking factor. The reactor power is multiplied by the hot spot factor to take into account the heterogeneous effect of the nuclear design model (i.e., power: +7%). Allocated rector power hot spot factor New model The hot spot factor related to the heterogeneous effect of the nuclear design model is eliminated by using detailed mesh. 16

18 2 nd Modification: Modification of Analysis Model Old model New model Horizontal 1/6 divided block model Each fuel rod model Vertical Fuel rod meshes 14 meshes Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006)

19 2 nd Modification: Modified Evaluation Method (1) Nuclear design code Fuel, control rod, core component, core internal structure design data Power density and neutron fluence distributions (2) FLOWNET In vessel thermal and hydraulic analysis code Coolant flow rate distribution (4) TEMDIM Fuel temperature analysis code Added (3) MVP Continuous energy Monte Carlo code Power distribution for each fuel rod Fuel temperature 18

20 2 nd Modification: Re evaluation Result of Fuel Temperature Old model (1 st modification) Fuel temperature New model (2 nd modification) Nominal temp. Maximum temp Old model New model Temp. difference Maximum fuel temp o C 1428 o C 35 o C Core average fuel temp o C 1018 o C 160 o C Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). 19

21 2 nd Modification: Re evaluation Result of Fuel Temperature Temperature distribution (Top) 1 Vertical position at a fuel column Temperature ( o C) (Bottom) Old model (1 st modification) Graphite block Coolant Sleeve outer surface Compact inner surface (nominal) (maximum) (Top) 1 Vertical position at a fuel column Temperature ( o C) (Bottom) New model (2 nd modification) Graphite block Coolant Sleeve outer surface Compact inner surface (nominal) (maximum) Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). 20

22 3 rd Modification: Re evaluation Result of Fuel Temperature Coolant temp. ( o C) Core inlet Analysis result (2 nd modification) Operation data at 950 o C operation in 2004 Center region Outside region Core outlet Center region Outside region Temperature rising Center region Outside region Maximum fuel temp. ( o C) 1,428 Analysis result using operation data (3 rd modification) 1,435 Ref: D. Tochio et al., Evaluation of Fuel Temperature on High Temperature Test Operation at the High Temperature Gas Cooled Reactor HTTR, Trans. At. Energy Soc. Japan, 5 [1], p.57 67(2006). 21

23 Fuel Temperature Measurement (Future Test) Objectives To develop the fuel temperature measuring technique for the prismatic HTGR To upgrade the core design technology for the prismatic HTGR Methods Temperature monitors using 22 kinds of melting wire in temperature range o C Irradiation performance of the melt wire temperature monitor has been confirmed by JMTR capsule irradiation test. Temperature monitors will be inserted into the fuel assembly to measure temperature distribution of the HTTR core. HTTR Melt wire temperature monitor Monitors will be stacked into hole under the dowel pin of the fuel assembly Fuel assembly Ref: Y. Tachibana et al., Test Plan using the HTTR for Commercialization of GTHTR300C, JAEA Technology (2009). 22

24 High Temperature Irradiation of the HTGR Fuel / Metallic FP Plate out Test (Future Test) Objectives To optimize the limitation of fuel failure under accident condition To investigate metallic FP (Sr, Cs, etc.) plateout behavior by using the real HTGR facility Methods A test fuel element loaded at the center column of HTTR and heated step by step up to 2000 o C Using temperature monitor to measure fuel temperature On line measurement of primary coolant radioactivity to estimate additional fuel failure fraction Plate out probe( ) will be settled in the primary circuit to measure metallic FP plateout concentration by PIE PIEs to investigate fuel failure and FP plateout Ref: Y. Tachibana et al., Test Plan using the HTTR for Commercialization of GTHTR300C, JAEA Technology (2009). Reactor to auxiliary cooling system HGC to SPWC Melt wire temp. monitor HGC Test fuel element IHX PPWC By-pass line For parallelloaded operation For singleloaded operation To pressurized water cooling system 23

25 Summary HTTR fuel temperature was re evaluated using the HTTR operating data and new analysis model. The summary of the re evaluation results of the maximum fuel temperature is the following: Design stage : 1492 o C 1 st modification using 850 C operation data : 1463 o C 2 nd modification new analysis model : 1428 o C 3 rd modification using 950 C operation data : 1435 o C We are planning to measure the HTTR fuel temperature using the melt wire temperature monitor. 24