WM’01 Conference, February 25-March 1, 2001, Tucson, AZ

RECOVERY AND TRANSMUTATION OF IODINE-129 IN AN ACCELERATOR-DRIVEN TRANSMUTATION SYSTEM

James C. Bresee U.S. Department of Energy

James J. Laidler Argonne National Laboratory,

Kimberly W. Thomas Los Alamos National Laboratory

ABSTRACT

Extensive repository system performance analyses have been carried out in the course of evaluating the feasibility of the proposed Yucca Mountain repository. These studies have shown that the primary contributors to public radiation dose as a consequence of release of radionuclides from the repository are isotopes of iodine, technetium and neptunium (129I, 99Tc, and 237Np). The United States Department of Energy is presently conducting a science-based R&D program, the Advanced Accelerator Applications Program, that addresses the technical issues associated with the transmutation of transuranic elements and long-lived fission products contained in the U.S. inventory of commercial Light Water Reactor (LWR) spent fuel. One of the main issues is the recovery and transmutation of iodine. The two fuels from which the iodine must be extracted are quite different, one (the LWR fuel) being an oxide fuel with zirconium alloy cladding, and the other most likely being a dispersion of metallic transuranics in a metallic zirconium matrix with stainless steel cladding. An aqueous solvent extraction process similar to the PUREX process will probably be used for LWR spent fuel processing, while a pyrochemical process seems to be most appropriate for the transmuter fuel. An overall system recovery target for iodine of 95% has been set as an initial goal of the program. Prior experience with iodine recovery in the PUREX process, both domestic and international, suggests that a satisfactory means can be developed for recovering the iodine contained in the LWR spent fuel. Iodine recovery in the less well-defined pyrochemical process for transmuter fuel appears to be feasible, but experimental verification and validation of the recovery steps is required. It is possible that the form of the recovered iodine from the two fuel types will be different, so it then remains to develop processes for incorporating the iodine in a standard target form, expected to be sodium iodide. Target design will also take into account means for accommodating the products of iodine and sodium transmutation.

INTRODUCTION

Transmutation by neutron irradiation has been studied for many years throughout the world and is currently under active investigation by the U.S. Department of Energy, within the Advanced Accelerator Applications (AAA) program of the Office of Nuclear WM’01 Conference, February 25-March 1, 2001, Tucson, AZ

Energy, Science and Technology. The purpose of this paper is to describe the AAA transmutation work as applied to 129I.

CURRENT U.S. HIGH-LEVEL NUCLEAR WASTE MANAGEMENT POLICY

The current high-level nuclear waste management program in the United States is based on policies established by the Nuclear Waste Policy Act of 1982, as amended. The Act states that the long-term management of such high-level waste will be through deep geologic disposal. In the case of civilian power reactors, the disposal form is to be untreated spent fuel suitably packaged. The Act also describes a rather intricate process for identification and selection of a disposal site, which through the end of calendar year 1987 produced an initial selection of nine possible sites; this was subsequently reduced to five and finally to three possible sites. In December 1987, the process was modified through an amendment of the 1982 law, resulting in the identification of only one of the candidate sites, Yucca Mountain, Nevada, for study or “characterization”. The Yucca Mountain site is composed of welded and unwelded volcanic ash, called “tuff” to a very great depth. It lies in a desert region on Federal land approximately 170 kilometers northwest of Las Vegas, at the western edge of the Nevada Test Site.

One characterization method has been “performance assessment”. The process involves the development of a series of mathematical models based on experimentation and natural analogues which describe the multistage process for movement of radioactivity from the disposal site to the accessible environment. In the case of Yucca Mountain, the accessible environment is defined as a region of marginal farming in Amargosa Valley some twenty to thirty miles southwest of the disposal site where agriculture based on irrigation is currently practiced by several thousand residents. Hydrologic studies have shown that ground water approximately 600 meters under Yucca Mountain flows very slowly to the southwest under Amargosa Valley and reaches the surface in Death Valley where it evaporates, depositing its dissolved minerals and salts in dry lake beds and other deposits. It is an important feature of the Yucca Mountain site that the regional hydrology is “closed”; that is, there is no connection to “open” systems like the Colorado River basin which reaches the ocean.

Assuming that Yucca Mountain were selected and licensed for disposal (a process which, if successful, will take about a decade), the multistage process for radioactivity release would include the following steps. First, rain would fall on the mountain, 95% of which would evaporate or run off. The remaining 5% would seep approximately 300 meters through the unsaturated zone until it encounters corrosion-resistant waste packages. Over geologic time (tens of thousands of years) such packages would begin to fail and soluble radionuclides would migrate into the unsaturated flow, eventually reaching the water table some 300 meters below. During that transport, some or most of the radionuclides would be absorbed on the underlying rock. The subsequent saturated flow to Death Valley would also involve absorption as well as dilution from rainfall along its path. Quite complex and integrated models allow a prediction, within bounding limits, of the potential exposure of Amargosa Valley residents to the kind and concentrations of radionuclides in future irrigation waters. WM’01 Conference, February 25-March 1, 2001, Tucson, AZ

A series of Total System Performance Assessments (or TSPA’s ) have been carried out for the Yucca Mountain site, in part to evaluate various repository design options but also to determine if the site can meet regulatory requirements. During 2001, the U.S. Department of Energy expects to issue a Site Recommendation Report for Yucca Mountain and an Environmental Impact Statement for the site, containing the latest TSPA information. Although not final at the time of this paper, the preliminary results show a continuation of the same trends that have been reported in previous TSPA’s. That is, from a potential human exposure standpoint, the most important radionuclides during the first 10,000 to 30,000 years of waste storage are 99Tc (213,000 year half- life) and 129I (15.7 million year half-life). Over the subsequent hundreds of thousands of years, 237Np (2.14 million year half-life) is the dominant radionuclide. The final approval of the Yucca Mountain site as a candidate for Nuclear Regulatory Commission licensing and the successful result of the licensing process will depend on whether the predicted performance of a Yucca Mountain repository is acceptable. The next few years will provide answers to that question.

TRANSMUTATION

In the future and as a possible improvement to current nuclear waste management practices, it is potentially possible to convert, through transmutation reactions, long-lived radionuclides such as radioiodine to short-lived or even stable isotopes. The process involves neutron irradiation, either in a nuclear reactor or in an accelerator-driven subcritical assembly. Long-lived actinides would be eliminated by fission, producing additional fission products for disposal or, if long-lived and mobile, for possible transmutation.

Most iodine fission products have short half-lives. After several years, the only remaining iodine in spent nuclear fuel will be stable 127I and very long-lived 129I. The probability of neutron capture by either nucleus (its capture cross section) is a function of neutron energy. Assuming the proper energy is attained, capture in either isotope would produce an unstable nucleus which would decay by beta emission to xenon. Stable 127I would be transformed to 128I, which would beta-decay with a 25-minute half-life to stable 128Xe; 129I would be transformed to 130I, which would beta-decay with a 12-hour half-life to stable 130Xe. Designing a program to collect fission product iodine, transform it to a suitable irradiation target, irradiate it with neutrons of the proper energy, chemically process the targets to recover untransmuted iodine for recycle and stable xenon for release, and repeat the process as needed is a challenging undertaking.

CHEMICAL PROCESSING OF SPENT FUEL AND THE RECOVERY OF IODINE FOR TRANSMUTATION

Each metric ton of LWR spent fuel, upon discharge from a reactor, contains almost 200 grams of 129I, together with another 45 grams of the stable iodine isotope 127I and about 10 grams of 131I. The 131I isotope poses the greater health concern, should it be released as for example in the case of a reactor accident; but in the case of fuel processing the short half-life (~8 days) of 131I means that this isotope has decayed away before fuel WM’01 Conference, February 25-March 1, 2001, Tucson, AZ processing begins. In the early days of nuclear fuel reprocessing, radioiodine was vented to the atmosphere, and in the rare events when fuel was processed after only very short cooling times, on the order of 15-20 days, the impact to the environment was significant (1).

In conventional aqueous solvent extraction reprocessing plants, a small fraction of gaseous iodine is released when the irradiated fuel rods are chopped prior to nitric acid leaching of the fuel. Upon dissolution, much of the iodine present in the fuel is oxidized to the elemental iodine state and appears as a gas; some is present as HI and some as HIO or HIO3. It is important to remove iodine from the dissolver solution before contacting this solution with the organic phase, because it will react with tributyl phosphate to form compounds that are difficult to remove later in the process. Removal of iodine from the dissolver solution is commonly done by sparging NOx gas through the solution in a slab evaporator/desorber to flush out the iodine. The flushed iodine is then sent to a caustic scrubber which retains most of the iodine, after which the effluent gas is sent through a HEPA filter and finally vented to the atmosphere. Other iodine sorption reagents tested successfully include fuming nitric acid and mercuric nitrate (2). Silver-impregnated zeolite molecular sieves are also used in some cases for iodine sequestration. Iodine compounds residing in the aqueous low-activity raffinate phase of the PUREX process are typically discharged to the sea. Airborne iodine effluents from the commercial reprocessing plants in France and the United Kingdom amounted to about 18 kg (about 120 GBq) from each plant over the period 1990-1995 (3), while discharges to the sea were nominally an order of magnitude greater (4).

For the treatment of spent LWR fuel in the AAA system, an aqueous process related to the PUREX process is envisioned. The process, which has been named UREX, differs from PUREX in that only the uranium constituent of the spent fuel is recovered as a purified stream, while the transuranics are confined to the aqueous raffinate without separation. The AAA system also differs in that it mandates the recovery of iodine for transmutation, because the radiotoxicity of 129I represents a considerable fraction of the projected release from the proposed Yucca Mountain geologic repository. With an overall system target of 95% iodine recovery, it will be necessary to have an efficient dissolver solution iodine desorber system, as well as a cell gas processing system to recover the iodine released during fuel rod chopping. These systems are reasonably technogically mature that the feasibility of recovering iodine for transmutation can be assured with a considerable degree of confidence. Figure 1 illustrates an iodine recovery system being considered for ATW applications; it is similar to systems currently in use in advanced reprocessing plants around the world.

But LWR fuel treatment is only part of the problem. The processing of LWR spent fuel has the purpose of recovering transuranic elements present in the fuel for fissioning in an accelerator-driven subcritical system, in addition to the recovery of long-lived fission product elements (iodine and technetium) for transmutation. The transuranics are to be incorporated in a non-fertile fuel form for irradiation in the transmuter system, and it is unlikely that a transmuter fuel can be designed to achieve 100% burnup of the transuranic elements present in the fuel. Therefore, the fuel must be recycled to extract the unburned WM’01 Conference, February 25-March 1, 2001, Tucson, AZ

TRU elements and newly-generated long-lived fission products. Because the fuel is expected to be recycled with short cooling times, and because the quantity of fuel to be processed is comparatively small, it has been decided that the fuel should be processed by pyrochemical means. This enables the treatment of fuels with quite high levels of decay heat generation, and is expected to result in a much lower quantity of high-level waste generation relative to aqueous processes. Furthermore, the dry pyrochemical processes envisioned permit operations with considerably higher masses of fissile materials while providing acceptable levels of protection against nuclear criticality events. All of these factors are important to establishing viable processing economics, but lead to processes for which the recovery of iodine has not been developed.

The nature of the pyrochemical process to be used for the treatment of ATW transmuter fuel is highly dependent on the composition and form of the fuel itself. At the present time, the prime candidate transmuter fuel is a dispersion of metallic transuranic elements in a metallic zirconium matrix. The pyrochemical process for treating this fuel is based on a chloride volatility step; a schematic flowsheet for the process is shown in Figure 2. A key presumption for iodine recovery in this process is that the radioiodine will be present in the highly reducing environment provided by the metallic fuel as an iodide of zirconium and certain fission product metals. With a lower vapor pressure than ZrCl4, the iodides will remain in the LiCl carrier salt while the zirconium matrix material is extracted as volatile ZrCl4. During the process of electrowinning of the TRU elements from the LiCl carrier salt, I- anions will react with chlorine gas liberated at the anode in an exchange reaction that produces gaseous iodine that is swept along by unreacted chlorine gas. It is then a matter of separating iodine from chlorine for incorporation in transmutation targets. Of course, it is possible that iodine will tend to be released at various other points in the process, and it will be necessary to determine that propensity by direct experimentation. Such tests are part of an intensive flowsheet validation experiment series that will be carried out over the next two to three years.

IODINE TRANSMUTATION TARGET PREPARATION

A target prepared from the iodine collected from spent LWR fuel will have significant components of both 129I, the radionuclide of interest for transmutation, as well as 127I, its stable counterpart which is also a product of the uranium fission process. The relative abundances of these isotopes is a function of the burnup of the original spent fuel, the addition of any iodine resulting from reprocessing of actinide transmutation targets, and any sources of stable iodine from the separations and preparation process. The presence of the stable isotope of iodine diminishes the transmutation efficiency for 129I because of competition for neutrons. Stable 127I will be converted to stable Xe through the transmutation process, but it will absorb neutrons in doing so.

Several important factors drive the selection of the transmutation target form. Ideally, the material would have a high melting point and low vapor pressure. If the target material is a compound rather than the pure element, the additional elements should have low neutron absorption cross sections to preserve neutron efficiency. The material needs tobe easy and safe to handle and lend itself to uncomplicated target preparation (eg, relatively WM’01 Conference, February 25-March 1, 2001, Tucson, AZ stable in air and moisture). A number of candidate target materials have been considered for iodine, but none meet all the above conditions in an ideal manner.

Elemental iodine has low melting/sublimation temperatures and thus very high vapor pressure issues, and its transmutation was given consideration by Logan, et al (5) and von Dincklange (6). Conti et al (7) and Konings et al (8) investigated possible iodide salt candidates and selected three for irradiation tests: NaI, PbI2, and CeI3. Samples of these materials were exposed to a neutron fluence on the order of 1.94x1026/m2 with a resulting iodine transmutation of 5-6% based upon the Xe produced. NaI was chosen as the most suitable, if not ideal, material. The lead compound showed evidence of melting and corrosion of the irradiation vial. The cerium compound posed handling problems. A compound yet to be experimentally tested would be 7LiI. The major advantage of this monoisotopic salt is that the neutron cross section for 7Li is smaller than that for 21Na. A major drawback is that the 7Li must be free of 6Li as the latter has a significant (n, a) cross-section that leads to the production of tritium. In all cases, release or containment of the Xe product must be addressed.

Regardless of the compound chosen, the transmutation process will produce a myriad of products in addition to the Xe decay products of the iodine transmutation (9). The cation will be transmuted in a competition for neutrons. There may be a buildup of cations due to iodine transmuting at a higher rate than that of the cation (desirable from a neutron conservation standpoint). This can lead to a variety of chemical reactions in the high pressure, high temperature environment during a long irradiation. For example, in the case of NaI as the target material, the formation of sodium metal must be considered.

Based upon the limited experimental data referenced above, NaI appears to be the most reasonable candidate for an iodine transmutation target. However, there are some extremely important issues yet to be resolved for transmutation applications. For example:

· will the exposure time for the NaI target be determined by the buildup of Xe gas, the capsule material, deleterious neutron effects of Xe buildup and the Na products, or some other phenomena? · do corrosive products form in the target capsules? Over what fluence/time? · does the NaI sinter and does this form retain Xe gas? · what other transmutation products are produced and does their buildup over long irradiations cause problems?

NUCLEAR CONSIDERATIONS FOR IODINE TRANSMUTATION

129 The radionuclide I is a beta emitter (Emax = 0.606 MeV) which also emits a weak gamma at 39.6 keV. Its fission yield from thermal fission of 235U is 0.75%. Stable 127I is also produced in uranium fission with a yield of 0.126%. Thus from prompt 235U fission, the iodine produced would be approximately 86% 129I and 14% stable 127I. Obviously this distribution would change with different fissile materials, if the neutron energy spectrum is altered, and if the residence time of the iodine in the actinide transmutation WM’01 Conference, February 25-March 1, 2001, Tucson, AZ fuel is increased. The latter reflects the in situ transmutation of the iodine fission products resulting from the transmutation and fissioning of actinides in the transmutation targets and is a function of the respective neutron capture cross sections for each nuclide.

In order for 129I transmutation to be successful, the rate of transmutation must be sufficient. The rate is dependent upon the neutron absorption cross section and the neutron flux. The thermal neutron cross sections for 129I are 20 and 10 barns to isomeric and ground states of 130I respectively. For 127I the cross section is 6.2 barns. The 127I (n, g) reaction results in 128I which decays to stable 128Xe and stable 128Te. Multiple (n, g) and (n, 2n) reactions (especially where harder neutron spectra are used) involving the starting materials and intermediate irradiation species must be considered in estimating the radiation fields during and post irradiation, for estimating neutron efficiency and iodine transmutation rates, and for ultimately defining the resulting waste stream from the iodine targets. In the event that competition for neutrons by the xenon products becomes a debilitating condition, continuous bleed-off of the xenon products could be considered. If 127I neutron absorption severely limits success, methods for the separation of 127I and 129I could be considered.

Obviously the various neutron absorption and other reaction cross section data are of paramount importance in assessing the feasibility of iodine transmutation. A thorough evaluation of the existing data on all potential nuclides that might be present in the targets should be undertaken to assess the need for additional measurements or calculations. The target nuclides should include both starting materials as well as those resulting from the various nuclear transformations during irradiation. Refined calculations for the transmutation of 129I must be done based upon the final blanket design.

REFERENCES

1. “Dissolving of Twenty-Day Metal at Hanford: Revision 1,” Report no. HW-17381- Del.-Rev. 1, 1950. 2. “Iodox Process tests in a Transuranium Element Campaign,” Report no. ORNL/TM- 6182, E. D. Collins and D. E. Benker, June 1979. 3. “Actinide and Fission Product Partitioning and Transmutation – Status and Assessment Report,” Nuclear Energy Agency, Organization for Economic Cooperation and Development, Paris, 1999, p. 208. 4. “Radiological Impacts of Spent Nuclear Fuel Management Options – A Comparative Study,” Nuclear Energy Agency, Organization for Economic Cooperation and Development, Paris, 2000, p. 33.

WM’01 Conference, February 25-March 1, 2001, Tucson, AZ

5. S. E. Logan, R. L. Conartz, H. S. Ng, L. J. Rahal, and Cl. G. Shirley, "Actinide Partitioning-Transmutation Program Final Report, VII. Long-Term Risk Analysis of the Geological Repository, " ORNL TM-6987, Oak Ridge National Laboratory (1980). 6. R-D. von Dincklange, "Apparatus for Transmutation of 129-I," Atomkernenergie- Kertechnik 38, 225-228 (1981). 7. A. Conti, J-P Ottaviani, R. J. M. Konings, and R. P. C. Schram, "Long-Lived Fission Product Transmutation Studies," Global'99, 7 (1999). 8. R. J. M. Konings, W. M. P. Franken, R. P. Conrad, J-F Gueugnon, J-C Spirlet, "Transmutation of Tc and I-Irradiation Tests in the Fram of the EFTTRA Cooperation, " Nucl. Technol. 117, 293-298, (197). 9. M. Attrep, Jr., "Accelerator Transmutation of 129-I, " LA-UR-92-64, Los Alamos National Laboratory, (1992).

WM’01 Conference, February 25-March 1, 2001, Tucson, AZ

Free Fission Gases Off-gas System

HNO3

Spent Chopped Dissolver Fuel Fuel Fuel Dissolver Solution Iodine Iodine Shear Desorber

Clarifier

Cladding Metal Hull Waste Form Cleanup 30%TBP in Kerosene Solvent Extraction Process

Stack

Silver- Caustic NO HEPA Zeolite Scrubber x Absorber Filter Column Column

Fig. 1. Schematic head-end flowsheet for the UREX process for treatment of LWR spent fuel preparatory to fissioning of the transuranic elements and transmutation of long-lived fission products, showing specifics of the iodine recovery process. WM’01 Conference, February 25-March 1, 2001, Tucson, AZ

Irradiated Fuel

Zr chloride; TRU chlorides; Sr, Cs chlorides; Chlorination RE chlorides; NMFP (metals) ZrCl4 Volatilization Reduction (in LiCl)

Cd Residual salt + cadmium Zr metal Cl2 Cd + Salt/Cadmium (to fuel fab.) NMFPs Separation Recycle to Cd Cadmium LiCl chlorination TRU chlorides step Distillation Sr, Cs chlorides RE chlorides Electro- NMFPs Iodine winning and Salt + FP chlorides Iodine Metal Other Extraction Waste Form NMFPs Technetium Separation Production TRU (metallic)

Ceramic Tc Fuel Waste Fabrication Form Target Fabrication Production

Fig. 2. Schematic flowsheet for pyrochemical process for treatment of zirconium-matrix/transuranic dispersion metallic fuel for the ATW subcritical transmuter.