DOE OSTI2012
In awareness of the risk caused by the proliferation of nuclear materials, the international community initiated in 1978 the Reduced Enrichment for Research and Test Reactors (RERTR) Program. The goal of the RERTR program is to seek solutions to convert the reactors using High Enriched Uranium (HEU) fuel ( 235 U/U ≥ 20 wt. %) to the use of Low Enriched Uranium (LEU) Fuel ( 235 U/U < 20 wt. %). Among the 200 reactors worldwide currently in the scope of the program, about 70 have been converted or shut down prior to conversion. In the U.S., six high reactors remain to be converted. One of the U.S. reactors is the High Flux Isotope Reactor (HFIR) located at Oak Ridge National Laboratory (ORNL). This multipurpose reactor is mainly used for neutron scattering experiments and isotope production. It has two fuel elements made of involute-shaped fuel plates. The ORNL Research Reactor Division (RRD) has prepared during FY2011 two series of calculations, one neutronic and one thermal-hydraulic (TH), in support of the conversion activities for HFIR. The neutronic calculations cover mainly the evaluation of the matrix of performance (cycle length and flux), the distribution of power in the fuel elements, kinetics parameters and reactivity coefficients. As defined by the RRD procedures, the calculations have to be reviewed. Per internal decision, the RRD proposed an external reviewer, Argonne National Laboratory (ANL). The results of the Argonne neutronic review are presented in this report. The results of the TH review are presented in a separate report. The main conclusion of the ORNL reactor physics analyses is that the proposed reference HFIR LEU core (using the so-called monolithic UMo fuel) could maintain the current level of performance (obtained with a HEU core at 85MW) if it is operated at 100MW. However, to operate within the same safety limits as the current HEU core, the thermal hydraulic analyses show that the peaking of the power distribution occurring at the bottom of the fuel plates has to be reduced. The RRD has found a potential solution by reducing the fuel thickness at the bottom of the fuel plate. This would require addition of an axial grading to the already radially-graded fuel. Thus, the review has been focused on independently recalculating all of the ORNL calculations and making an as fair as possible comparison between the current 85MW HEU core and the reference 100MW LEU core design. ORNL and ANL have used the same computational tools to perform the work. The codes MCNP and ORIGEN have been used to perform the steady state and depletion calculations, respectively. The communication between the two codes has been carried out by the code VESTA. Even though the code MCNP is able to handle complex geometries, it cannot model the involute shape. Nonetheless, ANL has developed an innovative methodology to model the involute shape with MCNP with an approximation that can be made as accurate as desired. Thus, ANL has been able to perform the calculations using an explicit (or discrete) description of the fuel elements. The current modeling used by ORNL consists in homogenizing the different fuel element components (cladding, fuel, coolant…). Although this method is easier to implement, it may also affect the reactor physics which may have an impact on the calculated results. Even though some differences in the physics parameters (flux for instance) have been detected between the HEU explicit (ANL) and homogenized (ORNL) models, they are small enough to not significantly impact the performance parameters. However, this is not the case for the LEU core where the differences in the physics parameters lead to significant differences in the performance parameters, especially the cycle length (found to be considerably longer in the explicit model). Nonetheless, the others parameters calculated by ANL (power distribution, kinetics parameters…) are in good agreement with those calculated by ORNL. Some inconsistencies have been detected in the original ORNL inputs that have been corrected in the reference ones used by ANL for the review and, therefore, the cycle length difference between the LEU homogenized and explicit models has been amplified. Consequently, the ANL comparison of the HFIR HEU and LEU cores has given different results than the ORNL results. The ORNL matrix of performance for the 100MW LEU reference design shows that the reactor cycle length would be as long and the flux as high as the current 85MW HEU core. In contrast, the ANL matrix of performance shows that the 100MW LEU design would have a significantly longer cycle length (~33days) than the current 85MW HEU core (~25days). In addition, it has been shown that, if the reactor is operated for 33days, the neutron fluence would be high enough to increase the overall reactor performance by a significant factor. If the reactor is forced to shut down after 25days, calculations show that the neutron flux would be high enough to maintain the performance of the neutron scattering experiments but losses are predicted to occur in several irradiation facilities. Increasing the cycle length may not be possible for other reasons. For instance, a longer cycle length can increase the cumulative oxide layer thickness growth on the plate. The increase in the neutron fluence may affect the lifetime of the reactor vessel and, by reducing the time for maintenance, pose operational challenges. An assessment of these as well as other potential questions associated with a 33 day cycle length is needed. Since such an assessment has not been carried out, ANL cannot conclude that the ORNL reference LEU core should or could operate at 100MW. The conclusions of the ANL review are different than those of the ORNL investigation mainly because the explicit and homogenized methodologies used to model the core are different. In this report, an attempt is made to justify and explain as clearly as possible why the ANL methodology is thought to be better suited for HFIR analyses. It is acknowledged that because the ANL and ORNL LEU models have not been benchmarked against experiments, it is possible that doubts could be raised concerning the ability of the models to correctly capture all of the physics behavior of the reactor.