technical feasibility of producing mo- 99 in triga 14 … feasibility of producing mo-99 in triga 14...

Institute for Nuclear Research

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

associated hot cellsC.Toma

Institute for Nuclear Research, Romania


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- -Production of 99Mo utilizing LEU targets in order to obtain the most used radioisotope for people health care in nuclear medicine - 99mTc.

-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 99Mo using the LEU –modified Cintichem process developed by Argonne National Laboratory.

OBJECTIVES


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-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.

The developing context


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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 99Mo in their generator and assure the dispensing of the product;


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 2006.

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REFERENCE TRIGA CORE

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TRIGA REACTOR

558,8 mm active length of the fuel is made up of erbium-uranium-zirconium hydride fuel moderator material (Er-U-ZrH1.6);

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

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TRIGA REACTOR FUEL PIN AND FUEL CLUSTER


IRRADIATION LOCATIONS IN TRIGA CORE

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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;

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

Irradiation capability


Reactor cores arrangement in the pool

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ACPR REACTORDRY

CAVITY

REACTOR TANK

RADIAL CHANNEL

TANGENTIAL CHANNEL REACTOR ENVELOPE

LOADING TUBE

SSR 14MW REACTOR

ROD CONTROL

(8)FUEL EXPERIMENTAL LOCATIONS FOR

LOOPS AND CAPSULES

BERILIUM REFLECTOR

EXPERIMENTAL LOCATIONS IN REFLECTOR

PLUG

TANGENTIAL CHANNEL

RADIAL CHANNEL

UNDERWATER NEUTRONOGRAPHY

THERMAL COLUMN STANDARD FOR NEUTRON FLUX CALIBRATION

NEUTRON DIFFRACTOMETER

PGNAA FACILITY

DRY NEUTRONOGRAPHY

SILICON DOPING CAVITIES


14 MW TRIGA Research Reactor

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14 MW TRIGA Research Reactor


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Description of INR Hot Cell FacilityHot 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.


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Description of INR Hot Cell Facility

nnected to the Examination Cell by a pneumatic rabbit post that allows the rapid transfer of the samples for metallography, 3 mm wall thickness. The Chemistry Cell has, also, one working station equipped with a SOVIS lead glass window and a pair of



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


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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).


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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;




Radiochemical cell modification


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-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;




Radiochemical cell modification


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Cell TypeInside

Dimensions

WxDxH[m]

Shielding Wall

Thickness[mm]

Atmo-sphere

Numberof

ShieldingWindows

MaximumActivity[TBq]

Functions

Alpha-GammaTransfer Cell

3.4x6x7 1100(Heavy Concrete

:3300

daN/m3)

Air 2 3.7x104- Loops and Capsules: Receiving,Disassembly-Reassembly andRod Extraction-Loading- Sheath Mechanical Property Tests: Tensile Test,Tube Burst Test- Radioactive Waste Treatment and Conditioning

Alpha-GammaExamination Cell

10x6x7 1100(Heavy Concrete

:3300

daN/m3)

Air orNit

rogen

7 3.7x104- Fuel Rod: Visual Inspection and Photography,Dimensional Measurements and TemperatureMeasurement, Eddy Current Testing,Gamma Scanning,Puncture and Fission Gas Measurement,Cutting and Resin Mounting, De-fuelling,Storage in the Pits


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Cell TypeInside

Dimensions

WxDxH[m]

Shielding Wall

Thickness[mm]

Atmo-sphere

Numberof

ShieldingWindows

MaximumActivity[TBq]

Functions

Alpha-GammaMetallography Cell

2x2x2 280(Steel)

Air 1 37 - Metallographic Specimen Preparation: Grinding,Polishing, Chemical and Electrochemical Etching

Alpha-GammaMicroscope Cell

1x1.2x0.8

200(Steel)

Air 1 3.7 - Optical Microscopy: Metallography, Ceramography,Image Analysis, Micro Hardness Test, Photography

Alpha-GammaChemistry Cell

2x2x2 280(Steel) plus 10 cm of lead

Air 1 37 - Target Dissolution and Chemical Preparationfor fission Molybdenum separation

Institute for Nuclear ResearchHot cell facility

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Post Irradiation Experiments Laboratory – LEPI


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


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New window and additional shielding to radiochemical cell


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Transfer cask mounted on transport vehicle


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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)


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Cross section through the irradiation device and target


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Irradiation device components


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Cutting tool for target disassembly and foil dissolver


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System for positioning, rotary motion and heating of the dissolver


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Foil dissolver


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Iodine recovery device


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Iodine and noble gases recovery device


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Iodine and noble gases recovery device


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Metallic supports (carriers) used in chemical processing


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Device for transferring LN2 into radiochemical cell;


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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).

Increasing irradiation capacityNew irradiation device

Institute for Nuclear ResearchNew irradiation device

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Institute for Nuclear ResearchData on target irradiation (1)

Target type: annular, LEU metallic foil between two Al cylinders;

Thermal neutron flux in the core center (perturbed): 1,65x1014 n/cm2xs;

Thermal neutron flux in the outward core center: up to 20% lower;

Reactor power: 10 MWIrradiation time: 5 days;Uranium mass per target: 16 g, 19.75 %

enrichment;40


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 U19, 75% enrichment;No. of max. irradiation channels: 5No of Curies per irradiation device at the end of

irradiation: 6640

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Data on target irradiation (3) Mo-99 recovery in chemical processing: 90%6-day Curies per irradiation device: 796The amount of 6-day Curies per 4 irradiation device

per week: 796x4=3184The amount of 6-day Curies per 4 irradiation device

per 11 month (yearly): 3184x4 x11=140 096The amount of Uranium mass to be irradiated in 11

month and 4 irradiation devices: 33792 g, 19.75 % enrichment;

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Calculation of 6-day Curies

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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)

1 192 796 8448 35024 7682 384 1592 16896 70048 15363 576 2388 25344 105072 23044 768 3184 33792 140096 30725 960 3980 42240 175120 3840

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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.


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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. 2000 g of low and medium level solid waste (plastics,

glassware, textures);-max. 300 ml of low and medium level liquid waste.


Radioactive waste treatmentTime evolution of fission products radioactive level after 5 irradiation days of 16 g U is:

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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)

74960 1222.8 426.6 131.4 52.4 26.2 15.4 3.3 1.7 1



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-147.

After 10 years cooling time main existing radioisotopes are: Kr-85, Sr-90, Y-90, Cs-137, Ba-137m, Pm-147

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Radioactive waste treatmentThe 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.

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No. of irradiation device U-mass/week U-mass/month U-mass/year

Liq. Waste/week (l)

Liq. Waste/year (l)

1 192 768 8448 2.4 105.62 384 1536 16896 4.8 211.23 576 2304 25344 7.2 316.84 768 3072 33792 9.6 422.45 960 3840 42240 12 528

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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.



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Radioactive waste system


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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


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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


Radioactive waste treatmentPreferred 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.

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Institute for Nuclear ResearchRadioactive waste: evaporation-

calcination

�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;

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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 299.1-85 “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”;

INTEGRATED MANAGEMENT SYSTEM


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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 AssuranceINR Quality Management System has also been authorized by CNCAN (Romanian Regulatory Body).

INTEGRATED MANAGEMENT SYSTEM


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INTEGRATED MANAGEMENT SYSTEMDocuments Control Records ControlProduct

Monitoring and Measurement

Determination and Review of

Requirements Related to the

Product

Planning of Product Realization

Purchasing

Resource Management

Financial- Bookkeeping Process

Communication

Monitoring and Measuring Devices Control

Software Control

Corrective/ Preventive Action

Nonconformities Control

Customer Satisfaction Monitoring

Processes Monitoring and Measurement

Integrated Management System

Planning

Management review

Internal Audit

Data Analysis

AllProcesses

AllProcesses

AllProcesses

AllProcesses

Basic Processes

Design Production and Service Provision Control

Research and Development(principal process)

Nuclear Installations Operation

Nuclear Installations Decommissioning

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Institute for Nuclear ResearchEnvironmental Management SystemThe requirements of the ISO 14001 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.

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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.

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Local transportThe 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.

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Mo-99 production logisticsWe 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.


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Mo-99 business logisticsA 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.


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ConclusionsThe 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 2013.

We consider our organization can commit a best effort schedule and technical approach for implementing a LEU based Mo-99 production process.

technical feasibility of producing medical isotopes, primarily Mo-99 in our TRIGA 14 MW reactor and associated hot cells.


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ConclusionsIn 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.


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ConclusionsMain 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.