127Xe and 37Ar Production in the University of Texas at Austin TRIGA Reactor

E. J. Artnak, R. Lester, S. Biegalski, D. Haas

The University of Texas at Austin, Nuclear Engineering Teaching Laboratory, 10100 Burnet Road, Building 159, Austin, TX 78758, Edward.Artnak@austin.utexas.edu

INTRODUCTION

The University of Texas at Austin has developed methods to produce isotopically pure noble gas samples for detector calibrations, quality control on detection systems, and environmental tracer studies. Methods were developed to create noble gas activities of 127Xe and 37Ar in excess of 1 Ci. These are produced through activation of 126Xe and 36Ar, respectively, in the 3-element irradiation facility within The University of Texas at Austin 1.1 MW TRIGA Mark II research reactor [1]. The methodology was specifically developed for this in-core irradiation facility. This work examines other methods that may be developed to increase the sample activities produced in similar activation experiments.

This summary contains a feasibility study for different irradiation methodologies that may be utilized to increase 127Xe and 37Ar sample activities. Further design studies would have to be conducted to assess a given experiment in more detail. A safety analysis of the facilities would also have to be conducted. FACILITY DESCRIPTION

The Nuclear Engineering Teaching Laboratory (NETL) has both in-core facilities for irradiations and beam port facilities for experiments. Neutron fluxes cover a wide range within the core, while beam port facilities are optimized for different experiments. In-Core Facilities

NETL’s in-core irradiation facilities are largely utilized for sample activation. The in-core facilities have the advantage of a higher neutron flux. However, sample geometry is often restricted in these facilities and their location within the core imposes some safety restraints. For the purposes of in-core gas irradiation, there are three primary in-core facilities of interest: 1) 3-element irradiator, 2) 7-element irradiator, and 3) central thimble. Table I outlines these in-core facilities along with their associated neutron fluxes.

The central thimble provides access to the highest neutron flux in the reactor. The central thimble is largely utilized to enhance activation or intensify radiation damage. Enhanced activation supports isotope production or neutron activation analysis, while heightened radiation damage supports material assessments involving high levels of

TABLE I. Neutron Flux for In-Core Irradiation Facilities

Facility (Flux values at 950 kW)

Thermal Neutron

Flux (n cm-2 s-1)

Total Neutron

Flux (n cm-2 s-1)

Maximum Sample Size

(Dia x H)

3-Element Irradiator 4 x 1012 2.8 x 1013 1.527ʺ x 15ʺ

7-Element Irradiator 2 x 1012 1.4 x 1013 2.25ʺ x 15ʺ

Central Thimble 1 x 1013 7.0 x 1013 1.25ʺ x 15ʺ

radiation exposure. For in-core irradiation facilities, the practicability of increasing the target gas pressure was examined. This increased atom density for a given irradiation volume results in an increased rate of radionuclide production. While the central thimble has a smaller working volume, it has a significantly larger neutron flux and the potential for the highest production rate of all in-core facilities. The evaluation of 127Xe and 37Ar production as a function of loading pressure in the central thimble was of significant interest in this work. Beam Port Facilities

The beam port facilities have a lower neutron flux in comparison to the in-core irradiation facilities. However, they include a larger space for irradiation vessels and allow increased flexibility for the design of experimental chambers. For beam port irradiations, the feasibility of irradiating liquid Ar and Xe was explored. The liquid phase of both Xe and Ar has a higher atom density, which facilitates increased activity production under the same irradiation conditions and geometry.

Some of the proposed design features are based on a similar cryogenic facility installed on beam port 3 for a cold neutron guide system. The primary components required for this facility to operate include the following: vacuum-jacketed flexible helium transfer line, cryorefrigerator cold head, helium compressor, water-cooled recirculating chiller system, beam port containment vessel for holding condensed gas during irradiation, and a large-volume reservoir tank (external to beam port) serving as a supply of target gas at ambient temperature and pressure. A possible facility design outlining these major components as integrated into the current TRIGA reactor core and beam port 1 (BP1) configuration is illustrated in Figure 1.

360

Transactions of the American Nuclear Society, Vol. 119, Orlando, Florida, November 11–15, 2018

Isotopes and Radiation: General—I

Fig. 1. Possible integration of cryogenic facility at BP1.

This facility provides the capability to condense 126Xe and 36Ar gas into a liquid safely in a closed system with minimal safety infrastructure requirements or concerns, especially over long irradiation times and without constant monitoring. The irradiation of 126Xe and 36Ar gas condensed into liquid state within the irradiation canister, which acts as a simple cold trap in this system, significantly increases the 126Xe and 36Ar atom densities within the irradiation canister resulting in the escalation of activity produced. A more detailed representation of proposed facility components and general interconnects is illustrated in Figure 2.

Fig. 2. Major components of BP1 cryogenic facility design. ACTIVITY CALCULATIONS

In order to calculate the estimated activity produced from a specified in-core or beam port (near-core) facility

irradiation, energy-distributed neutron flux profiles were determined using a previously established MCNPX model of the UT TRIGA reactor core [2, 3]. The flux was tallied in 63 groups with the same energy bin structure as the CINDER90 cross-section library. Using these cross-sections and energy-dependent flux profiles, the flux-weighted, energy-collapsed, neutron radiative capture cross-sections were calculated to be 2.26 b for 36Ar and 2.73 b for 126Xe. These cross-section values were assumed to be valid for total neutron fluxes (note: these are not thermal cross-sections) within the UT TRIGA reactor and near-core irradiation facilities.

The volume assumed for the central thimble was the maximum volume given by the dimensions in Table I. However, as the exact geometry and dimensions of the irradiation containment in both the central thimble and BP1 facility remained flexible (within reason), it was assumed here that the final design will be chosen such that self-shielding effects within the containment vessel are negligible. Furthermore, all calculations were performed assuming a 100-hour continuous irradiation at a reactor power of 950 kW.

Activity calculations for the central thimble were executed as a function of 36Ar and 126Xe gas pressure up to a maximum of 200 psi (absolute). The result was a linear response for activity produced versus containment loading pressure. The activity produced at maximum loading pressure (200 psi) was 35.2 Ci for 37Ar and 40.5 Ci for 127Xe. Although calculations were performed for containment pressures up to 200 psi, safety concerns over facility rupture and critical failure of nearby core infrastructure and fuel elements would likely limit the experimental pressure to a maximum range of 40 to 80 psi.

Activity calculations were also executed for the liquid phase of 36Ar and 126Xe within the near-core region of BP1. The total neutron flux utilized for the calculation was 7.0 x 1012 n cm-2 s-1. The activity was calculated as a function of cryogenic facility volume up to a maximum of 1,000 cm3. Again, the result was a linear response in activity produced versus volume of 36Ar and 126Xe liquid irradiated. The activity produced at maximum volume (1,000 cm3) was 800 Ci for 37Ar and 550 Ci for 127Xe. As a reference point, the beam port 3 cryogenic facility has a containment volume of 80 cm3, which may be considered a feasible volume. While the calculations were performed up to 1,000 cm3, this volume may be much higher than what is achievable for the given geometry and cooling requirements. The activity calculated for a containment volume of 80 cm3 was 64.1 Ci for 37Ar and 44.3 Ci for 127Xe. RESULTS

Increasing sample density through cryogenic condensation or deposition of gaseous samples increases activation rates. While, in-core facility calculations show that the development of a pressurized facility for the central

361

Transactions of the American Nuclear Society, Vol. 119, Orlando, Florida, November 11–15, 2018

Isotopes and Radiation: General—I

thimble would produce the highest activity per mole of target 36Ar or 126Xe, the pressure required to produce activity of the same order of magnitude as a reasonable 80 cm3 volume of liquid 36Ar or 126Xe (in a single run) on BP1 is unfeasible. Moreover, a highly specialized pressure vessel would need to be designed to contain the sample within the central thimble even at reasonably modest pressures of 40 to 80 psi. This effort would require a daunting safety analysis.

Alternatively, the increased densities of elements in liquid phase allow the use of a beam port facility despite the reduced flux. This also affords the option to utilize a larger containment volume in comparison to the in-core facilities. The calculations further show that a reasonable volume of 80 cm3 of liquid 36Ar or 126Xe will produce a 45-65x increase in activity compared to current production methods for the same irradiation time and reactor power. Significant effort will be required to design and build the cryogenic irradiation system on BP1. FUTURE WORK

The objective going forward is the design of a cryotrapping system for use in the existing BP1 facility to increase radioactive tracer production capability in a time- and cost-efficient manner. This design project must consider space limitations for the cold finger with reference to the effective flux region inside BP1. It must also evaluate the material properties of in-beam components, which are subject to neutron activation and radiation damage. Selection of materials with appropriate thermal properties is also crucial in order to limit the amount of heat transfer between system components and adjacent beam port infrastructure. Systems to introduce tracer gases into the system will also be designed to permit gas transfer without removal of the irradiation canister from BP1, while preventing over pressurization of the canister.

Preliminary testing will begin with smaller liquid nitrogen canisters and copper rods to simulate the expected results of the cryosystem on a scaled-down model. This will be achieved by sealing a partial length of copper rod in a liquid nitrogen Dewar and enclosing the remaining length within a vacuum-jacketed (cryotrapping) container. Natural argon and xenon gas will be introduced into the cryotrapping container from an external vessel at ambient temperature and pressure through an interconnecting, insulated pipe. The cryotrapping efficiency will be measured using different combinations of system components and characteristics such as interconnecting pipe length and diameter, external vessel volume, ambient temperature and pressure, carrier gas volume, thermal efficiency, as well as heat removal rate, among others.

Ultimately, these efforts intend to highlight critical characteristics and properties that largely govern the cryotrapping efficiency of the overall system and identify the most-effective selection of material components and

dimensions. Furthermore, this will also aid in the identification of any unforeseen issues that must be addressed in order for the full-scale system to be successful. REFERENCES 1. KB OLSEN, RR KIRKHAM, VT WOODS, DA

HAAS, JC HAYES, TW BOWYER, … SR BIEGALSKI, “Noble Gas Migration Experiment to Support the Detection of Underground Nuclear Explosions,” Journal of Radioanalytical and Nuclear Chemistry, 307, 3, pp. 2603-2610 (2016).

2. CM EGNATUK and SR BIEGALSKI, “Large Volume Gas Irradiation Procedure Developed at The University of Texas at Austin,” Journal of Radioanalytical and Nuclear Chemistry, 296, 1, pp. 223-226 (2013).

3. CM EGNATUK and SR BIEGALSKI, “Radioargon Production through the Irradiation of Calcium Oxide,” Journal of Radioanalytical and Nuclear Chemistry, 298, 1, pp. 475-479 (2013).

362

Transactions of the American Nuclear Society, Vol. 119, Orlando, Florida, November 11–15, 2018

Isotopes and Radiation: General—I