THE ORNL HIGH FLUX ISOTOPE REACTOR AND NEW ADVANCED FUEL TESTING CAPABILITIES J. L. MCDUFFEE and L. J. OTT Oak Ridge National Laboratory P.O. Box 2008, MS 6167 Oak Ridge, Tennessee 37831 Email:
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ABSTRACT The U.S. Department of Energy’s High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), was originally designed (in the 1960s) primarily as a part of the overall program to produce transuranic isotopes for use in the heavy-element research program of the United States. Today, the reactor is a highly versatile machine, producing medical and transuranic isotopes and performing materials test experimental irradiations and neutron-scattering experiments. The ability to test advanced fuels and cladding materials in a thermal neutron spectrum in the United States is limited, and a fast-spectrum irradiation facility does not currently exist in this country. The HFIR has a distinct advantage for consideration as a fuel/cladding irradiation facility because of the extremely high neutron fluxes that this reactor provides over the full thermal- to fast-neutron energy range. New test capabilities have been developed that will allow testing of advanced nuclear fuels and cladding materials in the HFIR under prototypic light-water reactor (LWR) and fast-reactor (FR) operating conditions.
1. Introduction—HFIR Description HFIR is a beryllium-reflected, pressurized, light-water-cooled, and moderated flux-trap-type reactor. The core consists of aluminum-clad involute-fuel plates, which currently utilizes highly enriched 235U fuel at a power level of 85 MWt. The reactor core, illustrated in Fig. 1, consists of two concentric annular regions, each approximately 61 cm in height. The flux trap is 12.7 cm in diameter, and the outer fueled region is 43.5 cm in diameter. The fuel region is surrounded by a beryllium annular reflector approximately 30.5 cm in thickness. The beryllium reflector is in turn backed up by a water reflector of effectively infinite thickness. In the axial direction, the reactor is reflected by water. The reactor core assembly is contained in a 2.44 m diameter pressure vessel, which is located in a 5.5 m cylindrical pool of water. Table 1 contains the characteristics of the vertical irradiation facilities in HFIR, which are shown as colored regions of Fig. 1. The facilities listed start with the target region inside the flux trap in the second column and proceed outward to the large Vertical Experiment Facility (VXF) positions (right-most column). Going from the target region to the large VXF position, the fast flux decreases by a factor of 100 and the thermal flux (without shields) decreases by a factor of 5. Also given in Table 1 are the characteristics for the RB* position with and without thermalneutron shields. Using the standard RB* Eu2O3 shield, the thermal neutron flux can be reduced by a factor of 50 without significantly affecting the fast flux, even after 8–10 cycles of operation. Thus, it is possible to tailor the neutron spectrum for specific experimental purposes and goals.
Fig. 1. Cross section through HFIR illustrating the primary experimental sites (left) and a picture of the reactor core (right)
Characteristic Fast flux, E > 0.1 MeV (1014 n/cm2·sec) Thermal flux, E < 0.1 MeV (1014 n/cm2·sec) Peak displacements per atom (dpa) per cycle, stainless steel Typical capsule diameter (mm) Number of available positions
Target 11 21 1.8 16 36
RB* unshielded 5.3 11 0.67 43 8
RB* with Eu2O3 shield
Small VXF
Large VXF
4.9
0.5
0.13
0.19
7.5
4.3
0.58 38 2
37 16
69 6
Tab 1: Characteristics of vertical HFIR irradiation facilities
2. LWR Experimental Design This project focused on the development of advanced fuel/cladding experimental capabilities in HFIR to replicate commercial LWR operating conditions (cladding and fuel temperatures, fuel average linear heat generation rates [LHGRs], and cladding fluence). Neutronics calculations indicated that the optimum experimental position in HFIR for the LWR experimental facility is the small vertical VXF locations in the beryllium reflector. The design of this experimental facility (illustrated in Fig. 2) allows simultaneous irradiation of nine simple uninstrumented capsules, each capsule containing a short fuel pin (fuel and cladding) with prototypic diametral dimensions. As shown in Fig. 2(a), the facility is a long cylindrical device with a thermal-neutron attenuation shield extending over the active length of the HFIR core (~61 cm in height). The cross section in Fig. 2(c) illustrates three cooling channels, each containing three stacked capsules [shown in Fig. 2(b)]. The capsules are cooled via primary coolant flow in the annulus between the capsule and the aluminum housing of the assembly. Provision is also made for three axial flux monitors (the only instrumentation in the package) that provide data for calibrating the neutronics models and confirming the safety basis for the facility. The neutron shield is cooled by primary coolant flow on the exterior of the assembly.
Fig. 2. VXF LWR experimental assembly
The sealed fuel pins [approximately 15 cm in height, Fig. 3(a)] contain up to 11 fuel pellets (enrichment 0.1 MeV) exceeding 1 1015 n/cm2s. To obtain the desired irradiation conditions, shielded rabbits must be employed. The initial experiments focus on the early stages of microstructural evolution in fuel materials, and the test articles utilized are standard transmission electron microscopy (TEM) disks, 3 mm in diameter and 200 m in thickness. The design (Fig. 4) of the TEM rabbit employs an internal gadolinium thermal-neutron shield that also acts as a specimen holder/spacer and as a heat conduit to maintain the specimen temperature within the desired range. Rabbit temperatures are monitored by SiC thermometry, and the neutron flux within the rabbit can be determined by replacing one (or more) of the TEM specimens with flux monitors. The reactivity of the shielded rabbit can be adjusted by adding/removing spacers. The first shielded rabbit experiments are expected to be conducted in the summer of 2011.
Aluminum housing
Steel holder SiC Thermometry Gadolinium
TEM Sample
Diffusion barrier
Fig. 4. Low-Fluence shielded TEM rabbit design
4. Summary The HFIR is a highly versatile test reactor, producing medical and transuranic isotopes and performing materials test experimental irradiations and neutron-scattering experiments. New test capabilities for the HFIR have been developed that will allow testing of advanced nuclear fuels and cladding materials under prototypic LWR and FR reactor operating conditions. An initial irradiation test of advanced LWR fuel/cladding specimens (supplied by commercial vendors) commenced in mid-2010, and FR experiments focused on fundamental fuel behavior are expected to start in mid-2011.
5. Acknowledgment This manuscript has been authored by UT-Battelle LLC under Contract No. DE-AC0500OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. This research has been sponsored by the Office of Fissile Materials Disposition, U.S. Department of Energy, National Nuclear Security Agency, under Contract DE-AC0500OR22725 with UT-Battelle, LLC.
