FBR and ATR fuel developments in JNC

Transcription

1 International Seminar on MOX Utilization February 19, 2002 FBR and ATR fuel developments in JNC Design and performance of MOX fuel in safety view points Plutonium Fuel Center, Tokai Works, Japan Nuclear Cycle Development Institute Tomoyuki Abe

2 CONTENTS 2 MOX fuel development in JNC Nuclear fuel cycle MOX fuel development FBR fuel development ATR fuel development Safety aspects of JNC MOX fuel What is MOX fuel? Design of MOX fuel Fuel inspection at reactor site Detail inspection of spent fuel Irradiation test in research reactor LWR-MOX fuel Structure design Irradiation condition Pu contribution to heat generation

3 Development of Nuclear Fuel Cycle 3 Mission of JNC is to develop nuclear fuel cycle to ensure national energy security Fast Breeder Reactor (FBR) and Advanced Thermal Reactor (ATR) have been a key issue. FBR, ATR MOX Fuel Spent Fuel Spent Fuel MOX Fuel Fabrication LWR Reprocessing Plutonium FBR and ATR fuel cycle

4 MOX Fuel Development Program 4 Fuel improvement has been continuously conducted through the cycle including fuel development, design, fabrication, irradiation, inspection, and evaluation Fabrication Design Development Driver Fuel Experimental Fuel MOX Fuel Technology Production Trial Manufacture Inspection P.I.E. Irradiation Test Irradiation P.I.E. Utilization

5 Development of FBR Fuel 5 FBR fuel development with JOYO and MONJU Plutonium production, full MOX core Accumulated experiences over 20 years in JOYO MONJU fuel development and fuel fabrication for initial core Plutonium recycling on a experimental scale Plutonium Fuel Production Facility 1960 History of FBR Fuel development FBR fuel development JOYO / MONJU / Demonstration Plant Experimental Fast Reactor JOYO and Fuel Monitering Facilities Fuel fabrication JOYO 525 S/As, MONJU 285 S/As Operation of JOYO Mk-I Mk-II Mk-III Operation of MONJU Reprocessing Laboratory Scale Proto-type FBR MONJU

6 Development of ATR Fuel 6 Development of ATR fuel with FUGEN Flexible utilization of plutonium Recycling experiences over 20 years with FUGEN Recycling of plutonium recovered from FUGEN MOX fuel Development for future ATR, burnup extension Plutonium Fuel Fabrication Facility History of ATR Fuel development ATR fuel development FUGEN / Demonstration Plant / Sum up Fuel fabrication 773 MOX S/As fabricated Proto-type ATR FUGEN Operation of FUGEN 726 MOX S/As loaded Reprocessing of MOX fuel Tokai Reprocessing Plant

7 What is MOX Fuel? 7 Characteristics of MOX fuel derive from differences between uranium and plutonium The difference is limited, most of MOX fuel technologies are common with those of UO 2 fuel Difference between Pu and U Technologies peculiar to MOX Neutronics Physical properties of mixed oxide Characteristics as radioactive isotope Common technologies between MOX and UO 2

8 Design of MOX Fuel 8 FBR fuel design All of driver fuels are MOX ATR fuel design Common design for both MOX and UO 2 Items peculiar to MOX Pu spot control Difference between Pu and U is considered in design calculation Fissile content is adjusted to be equivalent to UO 2 fuel Pu distribution of pellet Plutonium Spot Note Black spot corresponds Pu

9 Fuel Utilization & Inspection at Reactor Site 9 MOX driver fuel utilization FUGEN 726 S/As, JOYO 478 S/As Visual inspection and dimensional measurements of a great number of fuel S/As at pool-side in reactor site ensured that driver fuels had burnt successfully without any indication associated with cladding leakage Assembly Total Length Change mm FUGEN MOX Fuel FUGEN UO 2 Fuel Assembly Average burnup GWd/t Dimensional measurement of spent fuel at FUGEN

10 Post-Irradiation Exam of Driver Fuel 10 Some selected driver fuel S/As were transported to fuel inspection and examination facilities for Post-Irradiation Examination (PIE) PIE data obtained from various analyses, measurements and testings in hot-cells were utilized to assess in-pile fuel behavior and to check validity of fuel design Burning behavior was as expected in fuel design No indication associated with cladding leakage in various test results Fuel Assembly Observation Measurement Dismantling Fuel Rod Measurement Testing Cutting Data Data Pellet Sample Clad Sample Measurement Testing Data Measurement Testing Data Hot cell at OEC of JNC Typical flow chart of post-rradiation examination

11 Precise Irradiation Test in Research Reactors 11 Irradiation test in research reactors enables large variety of fuel testing for collecting essential data to ensure fuel safety in the reactor core Precise measurement with specially designed measuring devices Precise control of irradiation condition provides high quality data Comparison test Wrapper tube Compartment Test fuel rod B-type test fuel assembly used for irradiation test in JOYO

12 Severe Irradiation Test in Research Reactors 12 Irradiation test of special test fuel at severe condition Check of fuel specification In-pile fuel behavior during accident conditions Test Pellet FUGEN MOX pellet with plutonium spot artificially embedded on the surface PuO 2 artificially embedded Pu spot after irradiation No influence of the large Pu spot on integrity of cladding was observed. 0.9mm 1.1mm Test result Irradiation test to investigate effect of Pu spot on fuel behavior under accident condition

13 Structure Design of LWR, FBR & ATR Fuels 13 Structure and dimension of fuels Basic structure design of fuel pin is common, ATR design is similar to BWR design particularly Outer diameter of LWR fuel pins are between those of FBR and ATR FBR JOYO PWR ATR FUGEN Cross section of fuel rods BWR FBR PWR JOYO BWR ATR FUGEN Comparison of fuel structure design

14 In-core Conditions of LWR, FBR & ATR Fuels 14 In-core conditions Most of irradiation conditions of LWR MOX pellet are covered by irradiation experiences in FBR and ATR MOX Pu Content FUGEN LWR JOYO % Coolant Temperature Maximum Linear Rate Maximum S/A Burn-up FUGEN LWR JOYO JOYO LWR FUGEN kW/m FUGEN LWR JOYO Pin average GWd/t

15 Pu Contribution to In-core Heat Generation 15 LWR 100% UO2 Pu generation in UO2 fuel Pu contribution some 30% LWR/Pu-recycling MOX/core : 33% Pu contribution some 50% Experience in FUGEN MOX/core 34 72% Pu contribution % Experience in JOYO Enriched Uranium & Pu Pu contribution some 60% 100% BOC EOC BOC EOC BOC EOC BOC EOC 50% 0% U contribution to heat generation Pu contribution LWR / Once through 1 3MOX LWR / Pu-recycling 72 MOX 34 MOX FUGEN BOC:Beginning of operation Cycle EOC:End of operation Cycle Pu contribution to in-core heat generation by calculation 100 MOX JOYO

16 LWR-MOX Fuel & FBR/ATR Fuel 16 There are a great number of similarities among LWR, FBR & ATR fuels In-core irradiation conditions of LWR fuel is within those of FBR and ATR fuels Most of technologies on MOX are common among LWR, FBR and ATR Common technologies have been accumulated in JNC through development programs on FBR and ATR MOX fuels FBR Technology ATR Technology Common MOX Technology independent to reactor type Dependence to reactor type LWR Technology

17 Conclusion 17 MOX fuel development in JNC Technologies and experiences on MOX fuels have been accumulated in JNC through development programs of FBR and ATR. MOX fuel assemblies more than a thousand have been irradiated successfully at the experimental FBR JOYO and at the proto-type ATR FUGEN. Safety aspects of JNC MOX fuel Safety of JNC MOX fuel have been confirmed with a wide range of data on MOX fuels obtained from post-irradiation examination on JOYO and FUGEN driver fuels and from many irradiation experiments on various test fuels at research reactors. LWR-MOX Fuel Irradiation experience accumulated on FBR/ATR MOX fuel covers a wide range of in-core irradiation conditions which envelope those of LWR-MOX fuel. Most of fuel technologies for FBR and ATR are applicable to LWR-MOX.