Verification and Validation of TMAP7

The Tritium Migration Analysis Program, Version 7 (TMAP7) code is an update of TMAP4, an earlier version that was verified and validated in support of the International Thermonuclear Experimental Reactor (ITER) program and of the intermediate version TMAP2000. It has undergone several revisions. The current one includes radioactive decay, multiple trap capability, more realistic treatment of heteronuclear molecular formation at surfaces, processes that involve surface-only species, and a number of other improvements. Prior to code utilization, it needed to be verified and validated to ensure that the code is performing as it was intended and that its predictions are consistent with physical reality. To that end, the demonstration and comparison problems cited here show that the code results agree with analytical solutions for select problems where analytical solutions are straightforward or with results from other verified and validated codes, and that actual experimental results can be accurately replicated using reasonable models with this code. These results and their documentation in this report are necessary steps in the qualification of TMAP7 for its intended service

Laser Control of Dissipative Two-Exciton Dynamics in Molecular Aggregates

There are two types of two-photon transitions in molecular aggregates, that is, non-local excitations of two monomers and local double excitations to some higher excited intra-monomer electronic state. As a consequence of the inter-monomer Coulomb interaction these different excitation states are coupled to each other. Higher excited intra-monomer states are rather short-lived due to efficient internal conversion of electronic into vibrational energy. Combining both processes leads to the annihilation of an electronic excitation state, which is a major loss channel for establishing high excitation densities in molecular aggregates. Applying theoretical pulse optimization techniques to a Frenkel exciton model it is shown that the dynamics of two-exciton states in linear aggregates (dimer to tetramer) can be influenced by ultrafast shaped laser pulses. In particular, it is studied to what extent the decay of the two-exciton population by inter-band transitions can be transiently suppressed. Intra-band dynamics is described by a dissipative hierarchy equation approach, which takes into account strong exciton-vibrational coupling in the non-Markovian regime.Comment: revised version, fig. 8 ne

Verification and Validation of The Tritium Transport Code TMAP7

The TMAP Code was written at the Idaho National Engineering and Environmental Laboratory in the late 1980s as a tool for safety analysis of systems involving tritium. Since then it has been upgraded several times and has been used in numerous applications including experiments supporting fusion safety, predictions for advanced systems such as the International Thermonuclear Experimental Reactor (ITER), and estimates involving tritium production technologies. Its most recent upgrade to TMAP7 was accomplished in response to several needs. Prior versions had the capacity to deal with only a single trap for diffusing gaseous species in solid structures. TMAP7 includes up to three separate traps and up to 10 diffusing species. The original code had difficulty dealing with heteronuclear molecule formation such as HD and DT. That has been removed. Under pre-specified boundary enclosure conditions and solution-law dependent diffusion boundary conditions, such as Sieverts' law, TMAP7 automatically generates heteronuclear molecular partial pressures when solubilities and partial pressures of the homonuclear molecular species are provided for law-dependent diffusion boundary conditions. A further sophistication is the addition of non-diffusing surface species. Atoms such as oxygen or nitrogen or formation of hydroxyl radicals on metal surfaces are sometimes important in molecule formation with diffusing hydrogen isotopes but do not, themselves, diffuse appreciably in the material. TMAP7 will accommodate up to 30 such surface species, allowing the user to specify relationships between those surface concentrations and partial pressures of gaseous species above the surfaces or to form them dynamically by combining diffusion species or other surface species. Additionally, TMAP7 allows the user to include a surface binding energy and an adsorption barrier energy and includes asymmetrical diffusion between the surface sites and regular diffusion sites in the bulk. All of the previously existing features for heat transfer, flows between enclosures, and chemical reactions within the enclosures have been retained, but the allowed problem size and complexity have been increased to take advantage of the greater memory and speed available on modern computers. One additional feature unique to TMAP7 is radioactive decay for both trapped and mobile species. Recently, TMAP7 has undergone verification and validation processes to ensure its performance in a wide variety of problems. This paper describes the use and new capabilities of TMAP7 and presents results of verification and validation testing

Advanced Test Reactor LEU Fuel Conversion Feasibility Study -- 2006 Annual Report

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the U.S. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of these operating parameters, and the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U-235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U 235 enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U 235 loading in the LEU core, such that the differences in Keff and heat profile between the HEU and LEU core can be minimized for operation at 125 EFPD with a total core power of 115 MW. The Monte-Carlo coupled with ORIGEN2 (MCWO) depletion methodology was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the Keff versus EFPDs plot is similar in shape to the reference ATR HEU case. The LEU core conversion feasibility study can also be used to optimize the U-235 content of each fuel plate, so that the relative radial fission heat flux profile is bounded by the reference ATR HEU case. The detailed radial, axial, and azimuthal heat flux profiles of the HEU and optimized LEU cases have been investigated. However, to demonstrate that the LEU core fuel cycle performance can meet the UFSAR safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (OSCC, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores

Safety Analyses at the Idaho National Engineering and Environmental Laboratory Test Reactor Area - Past to Present

Test reactors are unique in that the core configuration may change with each operating interval. The process of safety analyses for test reactors at the Idaho National Engineering and Environmental Test Reactor Area has evolved as the computing capabilities, software, and regulatory requirements have changed. The evaluations for experiments and the reactor have moved from measurements in a set configuration and then application to other configurations with a relatively large error to modeling in three-dimensions and explicit analyses for each experiment and operating interval. This evolution is briefly discussed for the Test Reactor Area

3-D Thermal Evaluations for a Fueled Experiment in the Advanced Test Reactor

The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility. A test configurations has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates

Hardening Neutron Spectrum for Advanced Actinides Transmutation Experiments in the ATR

The most effective method for transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products is in a fast neutron spectrum reactor. In the absence of a fast rest reactor in the United States, initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. Such a test facility, with a spectrum similar but somewhat softer than that of the liquid-metal fast breeder reactor (LMFBR), has been constructed in the INEEL's Advanced Test Reactor (ATR). The radial fission power distribution of the actinide fuel pin, which is an important parameter in fission gas released modelling, needs to be accurately predicted and the hardened neturon spectrum in the ATR and the LMFBR fast neutron spectrum is compared. The comparison analyses in this study are peformed using MCWO, a well-developed tool that couples the Monte Carlo transport code MCNP with the isotope depletion and build-up code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations and detailed radial fission power profile calculations for a typical fast reactor (LMFBR) neutron spectrum and the hardened neturon spectrum test region in the ATR. The MCWO-calculated results indicate that the cadmium basket used in the advanced fuel test assembly in the ATR can effectively depress the linear heat generation rate in the experimental fuels and harden the neutron spectrum in the test region

Advanced Fuel Cycle Initiative - Projected Linear Heat Generation Rate and Burnup Calculations

This report provides documentation of the physics analysis performed to determine the linear heat generation rate (LHGR) and burnup calculations for the Advanced Fuel Cycle Initiative (AFCI) tests, AFC-1D, AFC-1H, and AFC-1G. The AFC-1D and AFC-1H tests consists of low-fertile metallic fuel compositions and the AFC-1G test consists of non-fertile and low-fertile nitride compositions. These tests will be irradiated in the East Flux Trap (EFT) positions E1, E2, and E3, respectively, during Advanced Test Reactor (ATR) Cycle 135B

Advanced Test Reactor LEU Fuel Conversion Feasibility Study (2006 Annual Report)

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of these operating parameters, and the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat profile between the HEU and LEU core can be minimized for operation at 125 EFPD with a total core power of 115 MW. The depletion methodology, Monte-Carlo coupled with ORIGEN2 (MCWO), was used to calculate K-eff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar in shape to the reference ATR HEU case. The LEU core conversion feasibility study can also be used to optimize the U-235 content of each fuel plate, so that the relative radial fission heat flux profile is bounded by the reference ATR HEU case. The detailed radial, axial, and azimuthal heat flux profiles of the HEU and optimized LEU cases have been investigated. However, to demonstrate that the LEU core fuel cycle performance can meet the UFSAR safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders (OSCCs), safety rods and regulating rod), and shutdown margins between the HEU and LEU cores

Neutronics and Thermal Hydraulics Study for Using a Low-Enriched Uranium Core in the Advanced Test Reactor -- 2008 Final Report

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis was performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff versus effective full power days (EFPDs) between the HEU and the LEU cores. The MCNP ATR 1/8th core model was used to optimize the U 235 loading in the LEU core, such that the differences in K-eff and heat flux profiles between the HEU and LEU cores were minimized. The depletion methodology MCWO was used to calculate K-eff versus EFPDs in this paper. The MCWO-calculated results for the LEU demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar to the ATR reference HEU case study. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, the proposed LEU (U-10Mo) core conversion case with nominal fuel meat thickness of 0.330 mm (13 mil) and U-235 enrichment of 19.7 wt% is used to optimize the radial heat flux profile by varying the fuel meat thickness from 0.191 mm (7.0 mil) to 0.330 mm (13.0 mil) at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19). A 0.8g of Boron-10, a burnable absorber, was added in the inner and outer plates to reduce the initial excess reactivity, and the peak to average ratio of the inner/outer heat flux more effectively. Because the B-10 (n,a) reaction will produce Helium-4 (He-4), which might degrade the LEU foil type fuel performance, an alternative absorber option is proposed. The proposed LEU case study will have 6.918 g of Cadmium (Cd) mixed with the LEU at the inner 4 fuel plates (1-4) and outer 4 fuel plates (16-19) as a burnable absorber to achieve peak to average ratios similar to those for the ATR reference HEU case study