Status of ISTC project K-1566

Transcription

1 The 7th Specialist Meeting on Recycling of Irradiated Beryllium October 22, 2012, Holiday Inn Executive Center, Columbia, MO, USA Status of ISTC project K Recycling Technology of Irradiated Beryllium - JPN Collaborator : K. TSUCHIYA (JAEA) Project Reader EU Collaborator : V. Kotov (IAE-NNC) : F. Druyts (SCK-CEN)

2 Research and Test Reactors in the World EU (~75) about 200 reactors in the world ASIA (~50) Beryllium is used in the research and testing reactors of 30% in the world. America (~60) Amount of Storage of used beryllium Japan : 3 ton World : 30~40 ton It is difficult to handle the used beryllium wastes. However, these wastes are produced on and on. :No use beryllium reflector :Use beryllium reflector Ref. 1

3 Outline of ISTC K-1566 Project It is important problem to recycle the used beryllium from the points of effective use of resources, reduction of radioactive waste and nuclear nonproliferation. Full Title : Recycling Technology of Irradiated Beryllium Project Period : Oct Mar Leading Institute : Institute of Atomic Energy (IAE) National Nuclear Center of The Republic of Kazakhstan (NNC-RK) Supporting Institutes : - NNC-RK/Institute of Atomic Energy (IAE) - Ulba Metallurgical Plant (UMP) Collaborators : - Japan Atomic Energy Agency (JAEA) - Studiecentrum voor Kernenergie - Centre d'étude de l'energie Nucléaire (SCK-CEN) 2

4 R&D Items in ISTC K-1566 Project This ISTC project consists nine tasks for R&D as follows, (1) Data study about beryllium technology and material purification from radioactive products, (2) Development of measuring methods of beryllium purification degree, (3) Register officially required documentation for conducting work with beryllium, tritium, chlorine, cobalt and irradiated beryllium transportation from Japan, (4) Development of laboratorial installation design and its equipment bundling, (5) Experiments conducting with un-irradiated beryllium on laboratorial installation and analysis of its results, (6) Experiments conducting with irradiated beryllium on laboratorial installation and analysis of its results, (7) Experiments on demonstration installation with non-irradiated beryllium, (8) Work conducting at demonstrational installation for purification and reprocessing of irradiated beryllium, (9) Development of generalized report. In March 2012, the final test was finished. 3

5 Treatment Procedure for Be Recycling Two methods to develop Be recycle technology were proposed. Wet method As the wet treatment, Be is initially converted to Be(BH 4 ) 2, (CH 3 ) 2 Be and BeCl 2 (dry method) --> [(CH 3 ) 3 C] 2 Be, then converted to BeH 2 by wet method. After that, metal-be can be obtained by thermal decomposition at high temperature; at 1500 C in Ar+H 2 (5%). Dry method Be is initially converted to BeCl 2 (volatile) by the reaction with Cl 2 (+ Ar) gas at 500ºC, then BeCl 2 is converted to metal-be powder at higher temperature (>1400ºC) by injection of H 2. 4

6 Beryllium Recycling Concept in ISTC Project The dry method is determined for beryllium recycling process because of reduction of radioactive waste and continual process. Basic flow of dry method are as follows. (1) Tritium release from irradiated beryllium (2) BeCl 2 production by the reaction between used beryllium and Cl 2 gas (BeCl 2 becomes gas at 482 C) Removal of Activated Impurity ( 60 Co, etc) Used Beryllium 500C Recovery of Beryllium 1400C (3) Be powder production by decomposition of BeCl 2 (Cl 2 gas is reused) 60 Co Separation (>96%) Cl 2 Beryllium Powder BeCl 2 (>99%) 5

7 Purification Results with un-irradiated Beryllium The conversion installation was prepared in the hood and the experiments were carried out with un-irradiated beryllium. Photograph of Device Results (for example) Initial Be Recovered Be Co Fe Al 0.73% 0.008% 0.038% 0.002% 0.15% 0.022% Conversion Flow Metal Be BeCl 2 Metal Be Cl 2 gas Measurement of impurities in metal beryllium Impurity content in the recovered beryllium were smaller than that of the initial beryllium. (Removal rate of Co : 99%) It was promising to purify impurity elements from irradiated beryllium. 6

8 Transportation of Irradiated Beryllium Transportation root, shield package, packing procedure were considered for transportation from JPN to KZ in this project. Institute of Atomic Energy (IAE) Transportation of irradiated beryllium by airplane (including of hermetic containers, shield containers and outer packages ) Preparation items Preparation of processing equipments for transportation of irradiated beryllium in the Japanese hot cell - Cutting device - Sealing containers From JAPAN Licensing and preparation of shield containers 7

9 Rate of Neutron Flux Irradiated Beryllium Samples from Japan Sample No. Length(mm) Center Weight(g) D E F G H I C J B Position K A Distance from Reactor Core (mm) A (Bq/g) B (Bq/g) C (Bq/g) D (Bq/g) E (Bq/g) F (Bq/g) G (Bq/g) H (Bq/g) I (Bq/g) J (Bq/g) K (Bq/g) 55 Fe Ni Co m Ag Be C T : Measuring nuclide : Non-measuring nuclide - Evaluation of Maximum Radioactivity at 1 March

10 Counts Count, imp. Results of Nuclides in Irradiated Beryllium(1) 434 kev Ag-108m 614 kev Ag-108m 662 kev Cs kev Ag-108m 1173 kev Co kev Co-60 g Spectrum of Irradiated Beryllium Nuclides in Irradiated Beryllium Ag-108m and Co-60 were measured by g spectrometer. Cs-137 was also observed. U content in beryllium? Ag-108m Ag-108m Cs-137 Ag-108m Co-60 Co-60 Half time (T 1/2 ) Decay Energy 55 Fe 2.7 y EC (100%) - 59 Ni y EC (100%) - 60 Co y b - (100%) 108m Ag 127 y EC (91.5%) IT(8.5%) MeV(g) MeV(g) MeV(g) MeV(g) MeV(g) 137 Cs 30 y b - (100%) MeV(g) 10 Be y b - (100%) 0.555MeV(b) 14 C 5730 y b - (100%) MeV(b) Energy (kev) Energy, kev 3 T 12.3 y b - (100%) MeV(b) 9

11 Results of Nuclides in Irradiated Beryllium(2) Flow rate (x10-12 mol/s) Tritium Release from Irradiated Beryllium 2.0 about 600 C 1280 C Temperature ( C) Elapsed Time (s) In the thermo-desorption measurements, tritium content was good agreement between the calculated value and the experimental data. 0 10

12 Evaluation of Nuclides in Experiments Comparison with calculated values and measurement values of each nuclide. Position A (Bq/kg) Position B (Bq/kg) Measuring Results (Bq/kg) Ratio between Calculation and Measurement A B M A/M B/M 55 Fe Ni Co m Ag Be C T Good agreement of tritium activity between calculated value and measured value Measurement method (Thermal desorption up to 1280 C) - Different of 60 Co and 108m Ag activities between calculated value and measured value? Shape effects during the measurements with the gamma-spectrometer Energy of 60 Co : 1.17MeV and 1.33MeV Energy of 108m Ag : 0.61MeV and 0.72MeV 11

13 Summary The tasks in this ISTC project were finished. Main results are as follows; (1)In the reaction between used-be and Chlorine(Cl 2 ) gas at 500ºC, BeCl 2 was generated, easily. (2)In the thermal-decomposition of BeCl 2, it was difficult to transform BeCl 2 into metal Be and Cl 2 gas because of high temperature operation. (3)In the preliminary demonstration test, high purity beryllium was recovered in this process. in future - Modification of the thermal-decomposition of BeCl 2 in the dry process - Proposal and necessity of new recycling process for large-scale 12