[en] A novel concept to detect pin-diversion from spent fuel assembly is proposed and described. The instrument will use multiple tiny neutron and gamma detectors in a form of cluster (detector cluster) and high precision driving system to collect radiation signatures inside pressurized water reactor (PWR) assembly. In order to validate our concept, a Monte Carlo study was done using a Monte Carlo code MCNP5. MONTEBURNS, a computational tool that links MCNP and ORIGEN, was used to produce accurate PWR spent fuel isotopic compositions. Monte Carlo simulations, using realistic fuel geometry and actual fuel material information, were performed to study radiation field inside a PWR spent fuel assembly. The preliminary Monte Carlo simulation study shows that indeed 2 dimensional neutron data, when obtained in the presence of missing pins, have data profiles distinctly different from the profiles obtained without missing pins

[en] A technical safeguards challenge has remained for decades for the IAEA to identify possible diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies. In fact, as modern nuclear power plants are pushed to higher power levels and longer fuel cycles, fuel failures (i.e., ''leakers'') as well as the corresponding fuel assembly repairs (i.e., ''reconstitutions'') are commonplace occurrences within the industry. Fuel vendors have performed hundreds of reconstitutions in the past two decades, thus, an evolved know-how and sophisticated tools exist to disassemble irradiated fuel assemblies and replace damaged pins with dummy stainless steel or other type rods. Various attempts have been made in the past two decades to develop a technology to identify a possible diversion of pin(s) and to determine whether some pins are missing or replaced with dummy or fresh fuel pins. However, to date, there are no safeguards instruments that can detect a possible pin diversion scenario to the requirements of the IAEA. The FORK detector system [1-2] can characterize spent fuel assemblies using operator declared data, but it is not sensitive enough to detect missing pins from spent fuel assemblies. Likewise, an emission computed tomography system [3] has been used to try to detect missing pins from a spent fuel assembly, which has shown some potential for identifying possible missing pins but this capability has not yet been fully demonstrated. The use of such a device in the future would not be envisaged, especially in an inexpensive, easy to handle setting for field applications. In this article, we describe a concept and ongoing research to help develop a new safeguards instrument for the detection of pin diversions in a PWR spent fuel assembly. The proposed instrument is based on one or more very thin radiation detectors that could be inserted within the guide tubes of a Pressurized Water Reactor (PWR) assembly. Ultimately, this work could lead to the development of a detector cluster and corresponding high-precision driving system to collect radiation signatures inside PWR spent fuel assemblies. The data obtained would provide the spatial distribution of the neutron and gamma flux fields within the spent fuel assembly, while the data analysis would be used to help identify missing or replaced pins. Monte Carlo simulations have been performed to help validate this concept using a realistic 17 x 17 PWR spent fuel assembly [4-5]. The initial results of this study show that neutron profile in the guide tubes, when obtained in the presence of missing pins, can be identifiably different from the profiles obtained without missing pins, Our latest simulations have focused upon a specific type of fission chamber that could be tested for this application

[en] Generally there are three methods of cross-section processing available when using the Scale computer code. They are NITAWL, BONAMI, and CENTRM, with CENTRM being the most common and accurate, but computationally expensive. In order to improve the accuracy of BONAMI (which uses the Bondarenko Method), new Bondarenko/F Factors were to be generated that will smooth out the current F Factors that are being generated using CENTRM. The case discussed here involves using a dedicated Monte Carlo code (MCNP5) to calculate the transport solution from which shielded cross-sections can be produced directly for the transport geometry/mesh. A simple program was created to parse and collect the tallied cross-sections from the MCNP output, which was fed into a modified CLAROL input (a module that replaces or adds data in an AMPX master library), which used the cross-sections obtained from MCNP to calculate new F-factors and update the SCALE library. This approach allows the use of other methods to generate the shielded cross-sections and for easy comparison to existing results. Initial proof-of-principle calculations were carried out for an various cases using various transport solvers, such as NEWT, KENO, and XSDRN, with BONAMI in SCALE. Poor results were obtained using cross-sections generated using infinite homogeneous cases, but good results were obtained by using pin cells in an infinite lattice. (authors)

[en] We suggest a new solution to the neutron slowing down equation in terms of multi-energy panels. Our motivation is to establish a computational benchmark featuring an ultra-fine group calculation, where the number of groups could be on the order of 100,000. While the CENTRM code of the SCALE code package has been shown to adequately treat this many groups, there is always a need for additional verification. The multi panel solution principle is simply to consider the slowing down region as sub regions of panels, with each panel a manageable number of groups, say 100. In this way, we reduce the enormity of dealing with the entire spectrum all at once by considering many smaller problems. We demonstrate the solution in the unresolved U3o8 resonance region. (authors)

[en] The High Flux Isotope Reactor (HFIR) currently uses highly enriched uranium (HEU) fabricated into involute-shaped fuel plates. It is desired that HFIR be able to use low enriched uranium (LEU) fuel while preserving the current performance capability for its diverse missions in material irradiation studies, isotope production, and the use of neutron beam lines for basic research. Preliminary neutronics and depletion simulations of HFIR with LEU fuel have arrived to feasible fuel loadings that maintain the neutronics performance of the reactor. This article illustrates preliminary models developed for the analysis of the thermal-hydraulic characteristics of the LEU core to ensure safe operation of the reactor. The beginning of life (BOL) LEU thermal flux profile has been modeled in RELAP5 to facilitate steady state simulation of the core cooling, and of anticipated and unanticipated transients. Steady state results are presented to validate the new thermal power profile inputs. A power ramp, slow depressurization at the outlet, and flow coast down transients are also evaluated. (authors)

[en] This study was carried out to assess the ability to eliminate meaningful quantities of americium in a primarily thermal neutron flux by 'spiking' modern BWR fuel with this minor actinide (MA). The studies carried out so far include the simulation of modern 10 x 10 BWR lattices employing the Westinghouse lattice physics code PHOENIX-4 alongside validation studies using MCNP5 models of the same lattices that were spatially depleted via the MONTEBURNS code coupling to ORIGEN. When considering the total inventory of minor actinides in Am-spiked pins, excluding isotopes of uranium and plutonium, the results indicate that a reduction of approximately 50% or more in the total mass inventory of these minor actinides is viable within the selected pins. Therefore, these preliminary results have encouraged the extension of this work to the development of improved lattice designs to help optimize the transmutation rates as well as absolute MA inventory reductions. The ultimate goal being to design batches of these advanced BWR bundles alongside multi-cycle core reload strategies. (authors)

[en] This work examined the impact of substituting FeCrAl cladding on a single PWR fuel rod and a single fuel assembly within a full core environment of standard Zircaloy cladded fuel. The goal was to provide insight on a pin-by-pin level of the power peaking effects caused by heterogeneity in the lattice, as well as to investigate some of the key fuel performance parameters. The neutronic analyses herein reported were performed using CASMO- 4/Simulate-3 models of a representative equilibrium PWR core with standard Westinghouse 17 x 17 assemblies, while the fuel performance analysis was performed using the Bison code with FeCrAl constitutive models currently under development at the University of Tennessee in collaboration with Oak Ridge National Laboratory. (Author)

[en] This article summarizes a few lessons learned in our early experiences in developing, delivering and implementing a series of distance learning classes for full-time undergraduate students enrolled in the combined-degree BS Mechanical + MS Nuclear Engineering 5-year and co-op based 'MNE- ACCEND' program at the Univ. of Cincinnati. This program is in its third year since inception and currently hosts approximately 35 undergraduate students enrolled in the graduating classes of 2008, 2009, and 2010, which is when these students are expected to complete their BS Mechanical and MS Nuclear Engineering degrees. In addition, 20+ newly confirmed students are expected to enter this program in the fall quarter of 2006 to become our Class of 2011. Therefore, the successful 'follow through' of the DL component of this program continues to be increasingly crucial as this student pipeline reaches a targeted steady-state of about 10 to 15 graduates per class. (authors)

[en] The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

[en] A machine learning process for determining cross-sections for use in nodal codes is being developed. This process foregoes the usual case matrix consisting of a nominal case of base conditions and both high and low branches in important state-point parameters to create orthogonal perturbations with an n-dimensional surface that does not require orthogonal branching. Cross-sections for a lattice are generated with CASMO-4 and given to the machine learning algorithm. A third degree ridge regression is performed to retain the same degree of accuracy as the standard nodal cross-section fit, but this regression includes cross terms. The cross-sections created by this fit agree well with the same state-point analyzed by CASMO-4. Differences between these cross-sections and those from traditional case matrices are discussed. Potential errors introduced with traditional orthogonal perturbations are explored. The effectiveness of these cross-sections is demonstrated by calculations of reactivity. (Author)