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Application of the INSTANT-HPS PN Transport Code to the C5G7 Benchmark Problem

Description: INSTANT is the INL's next generation neutron transport solver to support high-fidelity multi-physics reactor simulation INSTANT is in continuous development to extend its capability Code is designed to take full advantage of middle to large cluster (10-1000 processors) Code is designed to focus on method adaptation while also mesh adaptation will be possible. It utilizes the most modern computing techniques to generate a neutronics tool of full-core transport calculations for reactor analysis and design. It can perform calculations on unstructured 2D/3D triangular, hexagonal and Cartesian geometries. Calculations can be easily extended to more geometries because of the independent mesh framework coded with the model Fortran. This code has a multigroup solver with thermal rebalance and Chebyshev acceleration. It employs second-order PN and Hybrid Finite Element method (PNHFEM) discretization scheme. Three different in-group solvers - preconditioned Conjugate Gradient (CG) method, preconditioned Generalized Minimal Residual Method (GMRES) and Red-Black iteration - have been implemented and parallelized with the spatial domain decomposition in the code. The input is managed with extensible markup language (XML) format. 3D variables including the flux distributions are outputted into VTK files, which can be visualized by tools such as VisIt and ParaView. An extension of the code named INSTANTHPS provides the capability to perform 3D heterogeneous transport calculations within fuel pins. C5G7 is an OECD/NEA benchmark problem created to test the ability of modern deterministic transport methods and codes to treat reactor core problems without spatial homogenization. This benchmark problem had been widely analyzed with various code packages. In this transaction, results of the applying the INSTANT-HPS code to the C5G7 problem are summarized.
Date: June 1, 2011
Creator: Wang, Y.; Zhang, H.; Szilard, R. H. & Martineau, R. C.
Partner: UNT Libraries Government Documents Department

Thermally induced dispersion mechanisms for aluminum-based plate-type fuels under rapid transient energy deposition

Description: A thermally induced dispersion model was developed to analyze for dispersive potential and determine onset of fuel plate dispersion for Al-based research and test reactor fuels. Effect of rapid energy deposition in a fuel plate was simulated. Several data types for Al-based fuels tested in the Nuclear Safety Research Reactor in Japan and in the Transient Reactor Test in Idaho were reviewed. Analyses of experiments show that onset of fuel dispersion is linked to a sharp rise in predicted strain rate, which futher coincides with onset of Al vaporization. Analysis also shows that Al oxidation and exothermal chemical reaction between the fuel and Al can significantly affect the energy deposition characteristics, and therefore dispersion onset connected with Al vaporization, and affect onset of vaporization.
Date: December 1995
Creator: Georgevich, V.; Taleyarkham, R. P.; Navarro-Valenti, S. & Kim, S. H.
Partner: UNT Libraries Government Documents Department

A review of experiments and results from the transient reactor test (TREAT) facility.

Description: The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.
Date: July 28, 1998
Creator: Deitrich, L. W.
Partner: UNT Libraries Government Documents Department

Experimental capabilities of the transient reactor test (TREAT) facility.

Description: The TREAT facility was designed and built in the 1950s to provide a transient reactor for conducting safety experiments on reactor fuels. Throughout its almost 40-year history, it has proven to be a safe, reliable, and versatile facility, compiling a distinguished record of successful experiments. Several major improvements to the facility have been made, including an expansion of the building and of equipment handling capability, and enlargement of the access hole above the core, rearrangement of the reactor's control rods to provide more-uniform flux profiles, installation of improved reactor computer-control systems, a feedback system that safely allows real-time changes in power transients depending upon events occurring in the experiment, and several upgrades in the fast neutron hodoscope for improved experiment-fuel-motion diagnostics. The original TREAT fuel is still in use, however, since it appears to have no degradation from its many years of service.
Date: July 28, 1998
Creator: Crawford, D. C.
Partner: UNT Libraries Government Documents Department

Overview of Idaho National Laboratory's Hot Fuels Examination Facility

Description: The Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) of the Idaho National Laboratory was constructed in the 1960’s and opened for operation in the 1975 in support of the liquid metal fast breeder reactor research. Specifically the facility was designed to handle spent fuel and irradiated experiments from the Experimental Breeder Reactor EBRII, the Fast Flux Test Facility (FFTF), and the Transient Reactor Test Facility (TREAT). HFEF is a large alpha-gamma facility designed to remotely characterize highly radioactive materials. In the late 1980’s the facility also began support of the US DOE waste characterization including characterizing contact-handled transuranic (CH-TRU) waste. A description of the hot cell as well as some of its primary capabilities are discussed herein.
Date: September 1, 2007
Creator: Robinson, Adam B.; Lind, R. Paul & Wachs, Daniel M.
Partner: UNT Libraries Government Documents Department

Fast neutron hodoscope at TREAT: operational experience and improvements

Description: From nuclear science symposium; San Francisco, California, USA (14 Nov 1973). The multichannel hodoscope at TREAT is capable of detecting motion of fuel resulting from intentional meltdown of fuel pins contained within a test capsule. The system has been in operation in its present form since 1969, having participated in over 50 transients. Operational experience and evolutionary improvements are discussed. The rather complex hodoscope system, which is supported by an involved series of steps before results can be achieved, has worked well within design prediction. Computeraided reconstruction of the data has been provided to obtain displays, plots, and movies of fuel motion events. (auth)
Date: April 30, 1974
Creator: De Volpi, A.
Partner: UNT Libraries Government Documents Department

Performance studies of the parallel VIM code

Description: In this paper, the authors evaluate the performance of the parallel version of the VIM Monte Carlo code on the IBM SPx at the High Performance Computing Research Facility at ANL. Three test problems with contrasting computational characteristics were used to assess effects in performance. A statistical method for estimating the inefficiencies due to load imbalance and communication is also introduced. VIM is a large scale continuous energy Monte Carlo radiation transport program and was parallelized using history partitioning, the master/worker approach, and p4 message passing library. Dynamic load balancing is accomplished when the master processor assigns chunks of histories to workers that have completed a previously assigned task, accommodating variations in the lengths of histories, processor speeds, and worker loads. At the end of each batch (generation), the fission sites and tallies are sent from each worker to the master process, contributing to the parallel inefficiency. All communications are between master and workers, and are serial. The SPx is a scalable 128-node parallel supercomputer with high-performance Omega switches of 63 {micro}sec latency and 35 MBytes/sec bandwidth. For uniform and reproducible performance, they used only the 120 identical regular processors (IBM RS/6000) and excluded the remaining eight planet nodes, which may be loaded by other`s jobs.
Date: May 1, 1996
Creator: Shi, B. & Blomquist, R.N.
Partner: UNT Libraries Government Documents Department

Using the TREAT reactor in support of boron neutron capture therapy (BNCT) experiments: A feasibility analysis

Description: The technical feasibility of using the TREAT reactor facility for boron neutron capture therapy (BNCT) research was assessed. Using one-dimensional neutronics calculations, it was shown that the TREAT core neutron spectrum can be filtered to reduce the undesired radiation (contamination) dose per desired neutron more effectively than can the core spectra from two prominent candidate reactors. Using two-dimensional calculations, it was demonstrated that a non-optimized filter replacing the TREAT thermal column can yield a fluence of desired-energy neutrons more than twice as large as the fluence believed to be required and, at the same time, have a contamination dose per desired neutron almost as low as that from any other candidate facility. The time, effort and cost required to adapt TREAT for a mission supporting BNCT research would be modest.
Date: March 1, 1996
Creator: Grasseschi, G.L. & Schaefer, R.W.
Partner: UNT Libraries Government Documents Department

PFR/Treat Safety Experiments: HEDL Transient Test Program Engineering Test Plan

Description: The purpose of the PFR/TREAT Safety Test Program is to obtain experimental data of fuel pin behavior during hypothetical, unprotected accidents for cores of large liquid metal cooled fast breeder reactors. The steady state and transient experiments, which will be performed under the joint program, are to be as prototypic of fast reactor performance as is possible. The specific objectives of this document are: (1) dictate the activities and responsibilities for the HEDL Transient Test Program; (2) specify the technical requirements for the CO4, CO5, CO6 and CO7 test train (SPTTs); and (3) specify the technical requirement for the CO6 and CO7 Single Pin Test Loops (SPTLs). Specific requirements for single pin loop experiments beyond CO7 and multi pin experiments will be covered in the addenda to this test plan.
Date: March 1, 1981
Creator: Hoffman, M.A.; Metcalf, I.L. & Myron, D.L.
Partner: UNT Libraries Government Documents Department

Analytical and experimental investigation of the dispersion process during rapid transients for the aluminum-based nuclear fuel plates

Description: A thermally induced fuel-plate dispersion model was developed to analyze for dispersive potential and determine the onset of fuel plate dispersion for aluminum-based research and test reactor fuels. The effect of rapid energy deposition in a fuel plate was simulated. Several data types for aluminum-based fuels tested in the Nuclear Safety Research Reactor (NSRR) facility in Japan and in the Transient Reactor Test (TREAT) facility in Idaho, US, were reviewed. Analyses of experiments show that the onset of fuel dispersion is clearly linked to a sharp rise in the predicted strain rate, which further coincides with the onset of aluminum vaporization. Analysis also shows that aluminum oxidation and exothermal chemical reaction between the fuel and aluminum can significantly affect: the energy deposition characteristics and, therefore dispersion onset connected with aluminum vaporization, and the onset of aluminum vaporization.
Date: June 1, 1995
Creator: Georgevich, V.; Taleyarkhan, R.P.; Kim, S.H.; Fuketa, T.; Soyama, K. & Ishijima, K.
Partner: UNT Libraries Government Documents Department

Determination of the design excess reactivity for the TREAT Upgrade reactor

Description: The excess reactivity designed to be built into a reactor core is a primary determinant of the fissile loadings of the fuel rods in the core. For the TREAT Upgrade (TU) reactor the considerations that enter into the determination of the excess reactivity are different from those of conventional power reactors. The reactor is designed to operate in an adiabatic transient mode for reactor safety in-pile test programs. The primary constituent of the excess reactivity is the calculated reactivity required to perform the most demanding transient experiments. Because of the unavailability of supporting critical experiments for the core design, the uncertainty terms that add on to this basic constituent are rather large. The burnup effects in TU are negligible and no refueling is planned. In this paper the determination of the design excess reactivity of the TREAT Upgrade reactor is discussed.
Date: January 1, 1983
Creator: Bhattacharyya, S.K. & Hanan, N.A.
Partner: UNT Libraries Government Documents Department

TREAT neutron-radiography facility

Description: The TREAT reactor was built as a transient irradiation test reactor. By taking advantage of built-in system features, it was possible to add a neutron-radiography facility. This facility has been used over the years to radiograph a wide variety and large number of preirradiated fuel pins in many different configurations. Eight different specimen handling casks weighing up to 54.4 t (60 T) can be accommodated. Thermal, epithermal, and track-etch radiographs have been taken. Neutron-radiography service can be provided for specimens from other reactor facilities, and the capacity for storing preirradiated specimens also exists.
Date: January 1, 1981
Creator: Harrison, L.J.
Partner: UNT Libraries Government Documents Department

Simulation of the TREAT-Upgrade Automatic Reactor Control System

Description: This paper describes the design of the Automatic Reactor Control System (ARCS) for the Transient Reactor Test Facility (TREAT) Upgrade. A simulation was used to facilitate the ARCS design and to completely test and verify its operation before installation at the TREAT facility.
Date: January 1, 1984
Creator: Lipinski, W.C.; Kirsch, L.W. & Valente, A.D.
Partner: UNT Libraries Government Documents Department