Technical feasibility of producing Mo- 99 in TRIGA 14 MW reactor and associated hot cells

Transcription

1 Technical feasibility of producing Mo- 99 in TRIGA 14 MW reactor and associated hot cells C.Toma Institute for Nuclear Research, Romania 1

2 OBJECTIVES - -Production of 99 Mo utilizing LEU targets in order to obtain the most used radioisotope for people health care in nuclear medicine - 99m Tc. -To enhance the research reactor and other existing facilities utilization and obtain useful revenue. -To reduce nuclear proliferation concern, determined by HEU utilization by producing fission product 99 Mo using the LEU modified Cintichem process developed by Argonne National Laboratory. 2

3 The developing context -The Institute participates to IAEA CRP Developing techniques for small scale indigenous Molybdenum 99 production using Low Enriched Uranium fission or neutron activation together with other interested nuclear centers in developing technology for fission Molybdenum production. -Under this CRP, IAEA and ANL facilitates the transfer of knowledge for developing technology based on Cintichem modified process concerning target fabrication and chemical processing. 3

4 Motivation to develop fission Mo technology -the ability to manufacture target, using imported enriched Uranium metallic foils; -the ability to irradiate target in the existing TRIGA 14 MW reactor; -the ability to assure handling, disassembling and chemical processing of the target, in the existing hot cell facility taking into consideration some modifications and investments; -the ability of Institute for Nuclear Research to handle, treat and condition radioactive waste; -the existence in Romania, of the Center for Radioisotopes Production able to load 99 Mo in their generator and assure the dispensing of the product; 4

5 TRIGA REACTOR The initial TRIGA 14 MW active core was made up of 29 HEU fuel rod clusters enriched to 93% in U-235; The gradual conversion last many years due to TRIGA reactor ability to incorporate a great amount of excess reactivity and accordingly to require un-frequency re-fuelling; Complete conversion has been finished in May

6 REFERENCE TRIGA CORE 6

7 TRIGA REACTOR 558,8 mm active length of the fuel is made up of erbium-uranium-zirconium hydride fuel moderator material (Er-U-ZrH 1.6 ); Main differences between the two kinds of fuel are Uranium content and the enrichment which are respectively 10% and 93% in HEU fuel and 45% and 19.97% in LEU fuel 7

8 TRIGA REACTOR FUEL PIN AND FUEL CLUSTER 8

9 IRRADIATION LOCATIONS IN TRIGA CORE 9

10 Irradiation capability TRIGA core is flexible, different core configurations being easy obtained according to irradiation needs; Present configuration can be changed so that neutron flux in irradiation channels range in a narrow interval; The number of fuel clusters can be increased from 29 in present configuration up to 35 fuel clusters with congruent effects on neutron flux and irradiation locations; 10

11 Irradiation capability Adequate coolant flow for irradiation devices and even for a reactor power increasing; Reliable reactor operation; Other uses of reactor to reduce cost of 99 Mo. 11

12 Reactor cores arrangement in the pool DRY CAVITY ACPR REACTOR UNDERWATER NEUTRONOGRAPHY FUEL SSR 14MW REACTOR ROD CONTROL (8) EXPERIMENTAL LOCATIONS FOR LOOPS AND CAPSULES BERILIUM REFLECTOR PLUG LOADING TUBE THERMAL COLUMN STANDARD FOR NEUTRON FLUX CALIBRATION SILICON DOPING CAVITIES EXPERIMENTAL LOCATIONS IN REFLECTOR RADIAL CHANNEL REACTOR TANK TANGENTIAL CHANNEL REACTOR ENVELOPE TANGENTIAL CHANNEL RADIAL CHANNEL DRY NEUTRONOGRAPHY PGNAA FACILITY NEUTRON DIFFRACTOMETER 12

13 14 MW TRIGA Research Reactor 13

14 14 MW TRIGA Research Reactor 14

15 Description of INR Hot Cell Facility Hot cell facility is located adjacent to the TRIGA reactor and both are interconnected by an water transfer canal. Two large adjacent heavy concrete hot cells named Examination Cell and Transfer Cell, three adjacent steel hot cells named Metallography Cell, Microscope Cell and Chemistry Cell, operating area, control room, radiation control room, change room, truck lock, transfer canal area, service area and electric power supply system are located at first floor and equipped to perform effectively the PIE operations on reactor fuel and materials. 15

16 umatic rabbit post that allows the rapid transfer of the samples for metallography, Cell has, also, one working station equipped with a SOVIS lead glass window and a pair of Institute for Nuclear Research Description of INR Hot Cell Facility Chemistry Cell was constructed for the purpose of chemical processing operations for burnup measurement by mass spectrometry. The walls of the cell are 280 mm in thickness and can accommodate 37 TBq gamma activity. The inner stainless steel box of the cell has 2x2x2 m internal dimensions and 3 mm wall thickness. The Chemistry Cell has, also, one working station equipped with a SOVIS lead glass window and a pair of HWM-A100 master-slave manipulators. y 16

17 Target transfer to the Hot Cell Facility The water filled transfer channel provides a communication way between reactor tank and Hot Cell, facilitating underwater transfer of irradiated samples from core region directly into the receiving hot cell after 24 hours cooling time. Irradiated targets will be transferred and disassembled into the receiving cell. The irradiated foil will be transferred for chemical processing into radiochemical cell using an adequate container (PADIRAC system). 17

18 umatic rabbit post that allows the rapid transfer of the samples for metallography, Cell has, also, one working station equipped with a SOVIS lead glass window and a pair of Institute for Nuclear Research Radiochemical cell modification Focus was maintained on: -upgrading the existing hot cell and the associated systems in order to permit target processing through ANL procedures; -eliminate the interferences of the radiochemical processing cell with adjacent metallographic cell; -building up of additional required shielding to radiochemical processing cell to minimize dose; -modification of transfer cask capable of supplying the required shielding of 9 KW target during target transportation from disassembling cell to radiochemical processing cell; 18

19 umatic rabbit post that allows the rapid transfer of the samples for metallography, Cell has, also, one working station equipped with a SOVIS lead glass window and a pair of Institute for Nuclear Research Radiochemical cell modification -ensuring electric drive of the access door opening and closing of the radiochemical cell; -design and fabrication of transfer system of final product from radiochemical processing cell into the shipping container; 19

20 Cell Type Inside Dimensio ns WxDxH [m] Shielding Wall Thickness [mm] Atmosphere Number of Shielding Windows Maximum Activity [TBq] Functions Alpha-Gamma Transfer Cell Alpha-Gamma Examination Cell 3.4x6x (Heavy Concrete : 3300 dan/m 3 ) 10x6x (Heavy Concrete : 3300 dan/m 3 ) - Loops and Capsules: Receiving, Air 2 3.7x10 4 Disassembly-Reassembly and Rod Extraction-Loading - Sheath Mechanical Property Tests: Tensile Test, Tube Burst Test - Radioactive Waste Treatment and Conditioning Air or Nit rogen 7 3.7x Fuel Rod: Visual Inspection and Photography, Dimensional Measurements and Temperature Measurement, Eddy Current Testing, Gamma Scanning, Puncture and Fission Gas Measurement, Cutting and Resin Mounting, Defuelling, Storage in the Pits 20

21 Cell Type Inside Dimensio ns WxDxH [m] Shielding Wall Thickness [mm] Atmosphere Number of Shielding Windows Maximum Activity [TBq] Functions Alpha-Gamma Metallography Cell Alpha-Gamma Microscope Cell 2x2x2 280 (Steel) 1x1.2x (Steel) Air Metallographic Specimen Preparation: Grinding, Polishing, Chemical and Electrochemical Etching Air Optical Microscopy: Metallography, Ceramography, Image Analysis, Micro Hardness Test, Photography Alpha-Gamma Chemistry Cell 2x2x2 280 (Steel) plus 10 cm of lead Air Target Dissolution and Chemical Preparation for fission Molybdenum separation 21

22 Hot cell facility 22

23 Post Irradiation Experiments Laboratory LEPI 23

24 Rear wall of radiochemical cell Electric drive of the access door opening and closing 24

25 New window and additional shielding to radiochemical cell 25

26 Transfer cask mounted on transport vehicle 26

27 Implementing a LEU based Mo-99 production process. Equipments design and fabrication for small scale production All equipments for target fabrication, irradiation, disassembling, chemical processing were designed and fabricated: -the equipment for noble gases and iodine recovery; -the system for positioning, rotary motion and heating of the dissolver during dissolving process; -the metallic supports (carriers) used in chemical processing during separation and purification; -the device for transferring LN2 into radiochemical cell; -also, evaluation of equipment performance, reliability and adequacy of design determined us to improve some of these (target cutting tool and system for positioning, rotary motion and heating of the dissolver) 27

28 Cross section through the irradiation device and target 28

29 Irradiation device components 29

30 Cutting tool for target disassembly and foil dissolver 30

31 System for positioning, rotary motion and heating of the dissolver 31

32 Foil dissolver 32

33 Iodine recovery device 33

34 Iodine and noble gases recovery device 34

35 Iodine and noble gases recovery device 35

36 Metallic supports (carriers) used in chemical processing 36

37 Device for transferring LN2 into radiochemical cell; 37

38 Increasing irradiation capacity New irradiation device To increase irradiation channel capacity a new irradiation device was design having 4 target holders; Each target holder can support 3 targets; Each target can contain 2 Uranium foil weighing 16 g; Total Uranium mass accepted by new irradiation device is 192 g (3x16x4). 38

39 New irradiation device 39

40 Data on target irradiation (1) Target type: annular, LEU metallic foil between two Al cylinders; Thermal neutron flux in the core center (perturbed): 1,65x10 14 n/cm 2 xs; Thermal neutron flux in the outward core center: up to 20% lower; Reactor power: 10 MW Irradiation time: 5 days; Uranium mass per target: 16 g, % enrichment; 40

41 Data on target irradiation (2) Target power: 1 Kw /g U 19, 75% enrichment; Molybdenum production: 40 Ci Mo-99 / g U 19, 75% enrichment; Irradiation device capacity: 12 targets, 192 g U 19, 75% enrichment; No. of max. irradiation channels: 5 No of Curies per irradiation device at the end of irradiation:

42 Data on target irradiation (3) Mo-99 recovery in chemical processing: 90% 6-day Curies per irradiation device: 796 The amount of 6-day Curies per 4 irradiation device per week: 796x4=3184 The amount of 6-day Curies per 4 irradiation device per 11 month (yearly): 3184x4 x11= The amount of Uranium mass to be irradiated in 11 month and 4 irradiation devices: g, % enrichment; 42

43 Calculation of 6-day Curies 43

44 Data on Mo-99 production No. of irradiation device U-mass/week 6-day Ci/week U-mass/year 6-day Ci/year U-mass/month(g)

45 Chemical processing capacity -Only one radiochemical cell is available for the time being for chemical processing of 16 g U target and using existing equipment; -In 24 hours could be processed 3 similar targets, hence 48g of irradiated Uranium; -In fact, should be necessary to process in a cell in 24 hours at least the content of an irradiation device occupying an irradiation location (192 g U); -New processing equipment have to be design, fabricated and tested; -Large scale Mo-99 production would need a greater number of radiochemical cells, maybe 3 or 4. 45

46 Radioactive waste treatment Without taking into consideration Uranium recovery, the estimation for average amounts of radioactive waste for a target containing 16 g of 19.75% enriched Uranium are: -max. 200 ml high rad. level of liquid ; -max. 200 g of metallic waste; -max g of low and medium level solid waste (plastics, glassware, textures); -max. 300 ml of low and medium level liquid waste. 46

47 Radioactive waste treatment Time evolution of fission products radioactive level after 5 irradiation days of 16 g U is: 0 d (Ci) 10 d (Ci) 30 d (Ci) 90 d (Ci) 180 d (Ci) 270 d (Ci) 1 y (Ci) 3 y (Ci) 5 y (Ci) 10 y (Ci)

48 Radioactive waste treatment After 5 years cooling time main existing radioisotopes are: Kr-85, Sr-90, Y-90, Ru-106, Rh-106, Cs-137, Ba-137m, Ce-144, Pr-144, Pm After 10 years cooling time main existing radioisotopes are: Kr-85, Sr-90, Y-90, Cs-137, Ba-137m, Pm

49 Radioactive waste treatment The next table contains data on Uranium quantities to be irradiated and liquid waste produced during chemical processing as function of number of irradiation device used for Mo-99 production; About 0.5 cubic meter of high radioactive liquid waste will be produced each year by chemical processing of the content of 4 irradiation devices; This amount have to be stored between 5 and 10 years. 49

50 Radioactive waste treatment No. of irradiation device U-mass/week U-mass/month U-mass/year Liq. Waste/week (l) Liq. Waste/year (l)

51 Radioactive waste treatment The treatment and elimination of gaseous waste is solved by an efficient off gas system. Xenon and Iodine isotopes are trapped on special materials for decay before elimination in atmosphere. Volume reduction of solid waste is carried out through cutting and subsequently, they are conditioned in special containers. Liquids containing enriched Uranium will be stored till a decision concerning recycling for Uranium recovery will be taken. 51

52 Radioactive waste system 52

53 10 CFR Part 20 Dose Standards 2 millirems in any one hour from external sources in an unrestricted area; 100 millirems in a calendar year (sum of external and internal radiation) in a controlled or unrestricted area 53

54 10 CFR Part 50 Design Objectives Liquids 3 millirems/year to the whole body 10 millirems/year to any organ Gases 5 millirems/year to the whole body 15 millirems/year to the skin Solids and Iodine15 millirems/year to any organ 54

55 Radioactive waste treatment Preferred option would be the transfer of these waste to a recycling center. If not possible after 5 to10 years of cooling time, Uranium from liquid waste could be extracted and remaining waste treated and conditioned for storage in National Storage for Radioactive Waste (a former Uranium mine) Other waste solutions, containing fission products and actinides will be mixed and adjusted to neutral ph and immobilized with cement. 55

56 Radioactive waste: evaporationcalcination Liquid radioactive waste for long-term storage must be subject to treatment. This means (as in original HEU Cintichem process): -volume reduction by evaporation; -solid generation by calcination; 56

57 INTEGRATED MANAGEMENT SYSTEM INR has developed and implemented an Integrated Management System in accordance with the following documents requirements: ISO 9001:2000 Quality Management Systems-Requirements ; ISO 14001:2004 Environmental management systems. Requirements with guidance for use ; OHSAS 18001:1999 Occupational health hand safety management systems. Specification ; ISO 17025:2005 General requirements for the competence of testing and calibration laboratories ; ASME code, 1980 edition, without Addendum, Section III, NCA 4000; CAN 3 Z Quality Assurance Program Category 1 ; IAEA Safety Standards, GS-R-3, The Management System for Facilities and Activities ; IAEA Safety Standards, GS-G-3.1, Application of the Management System for Facilities and Activities ; NMC-01:2003 Norms concerning authorization of quality management systems applied to construction, operating and decommissioning of nuclear installations ; NMC-02:2003 Norms concerning general requirements for quality management systems applied to construction, operating and decommissioning of nuclear installations ; 57

58 INTEGRATED MANAGEMENT SYSTEM NMC-04:2003 Norms concerning specific requirements for quality management systems applied to research and development activities for nuclear field ; NMC-05:2003 Norms concerning specific requirements for quality management systems applied to nuclear installations design ; NMC-07:2003 Norms concerning specific requirements for quality management systems applied to manufacturing of products for nuclear installations ; NMC-10:2003 Norms concerning specific requirements for quality management systems applied to nuclear installations operating ; NMC-11:2003 Norms concerning specific requirements for quality management systems applied to nuclear installations decommissioning ; NMC-12:2003 Norms concerning specific requirements for quality management systems applied to design and use of software for research, design, analysis and calculations intended to nuclear installations. INR Quality Management System has been certified by Lloyd s Register Quality Assurance INR Quality Management System has also been authorized by CNCAN (Romanian Regulatory Body). 58

59 INTEGRATED MANAGEMENT SYSTEM Determination and Review of Requirements Related to the Product Documents Control Basic Processes Product Monitoring and Measurement Records Control Processes Monitoring and Measurement All Processes Planning of Product Realization Purchasing Resource Management Design Production and Service Provision Control Research and Development (principal process) Integrated Management System Planning Management review All Processes Data Analysis All Processes Financial- Bookkeeping Process Nuclear Installations Operation Nuclear Installations Decommissioning Internal Audit Communication All Processes Monitoring and Measuring Devices Control Software Control Corrective/ Preventive Action Nonconformities Control Customer Satisfaction Monitoring 59

60 Environmental Management System The requirements of the ISO EMS Standard are followed by the Institute Management Division s managed environmental program. These requirements include: committing to continual improvements of its environmental performance; complying with all relevant legislation; committing to the prevention of pollution, and ensuring that the adverse environmental impacts of its activities, products and services are as low as reasonably possible. 60

61 Environmental Management System Annual reviews and evaluations maintain the performance of the EMS by the Environmental Management Review Team which determines the progress of objectives and targets for the Significant Environmental Aspects and establishes new targets and objectives as needed. 61

62 Local transport The site of institute is situated at 130 km north of Bucharest (the capital) in Mioveni, a small town near Pitesti in the Arges county; There is a high way between Pitesti and international airport of Bucharest; The quality of roads are acceptable and the transportation time needed from institute to the airport is around 2.5 hours. 62

63 Mo-99 production logistics We have to confront with two kind of logistics: -production logistics for Mo-99; -Mo-99 business logistics; To coordinate a sequence of resources to carry out some project or some product is usually done in our institute; Track and tracing, which is also an essential part of production logistics - due to product safety and product reliability issues is not a familiar subject because we did not produce any isotopes for medical use. 63

64 Mo-99 business logistics A supply chain in moving a product fabricated in our institute to different customers was not the case up to now. But the institute is able to be part of a system of organizations, people, technology, activities, information and resources involved in a functional supply chain. Having the right item in the right quantity at the right time at the right place for the right price in the right condition to the right customer is a goal to be achieved. 64

65 technical feasibility of producing medical isotopes, primarily Mo-99 in our TRIGA 14 MW reactor and associated hot cells. Institute for Nuclear Research Conclusions The response regarding future production of Mo-99 to inquires from U.S. Department of Energy considers technical feasibility as being possible in our TRIGA 14 MW reactor and associated hot cells. We affirm our organization s willingness and ability to commit to producing Mo-99 for the international market if needed using LEU target by October We consider our organization can commit a best effort schedule and technical approach for implementing a LEU based Mo-99 production process. 65

66 Conclusions In formulating this response we assumed an appropriate collaboration with different experts and organizations for the next two -three years to help to upgrade the process and equipments to a large production capacity to meet the above presented irradiation capacity. The availability of future partners to work with the INR is considered to be determinant for the action success. 66

67 Conclusions Main issues to be addressed -LEU material and foils supply; -Development of chemical processing capacity in our institute considering actual assets and new investments from Romanian funds; -Future TRIGA LEU fuel supply; -Radioactive waste-including LEU recovery management; -Transportation issues. 67