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Assessment of the SE2-ANL code using EBR-II temperature measurements

Description: The SE2-ANL code is a modified version of the SUPERENERGY-2 code [1]. This code is used at Argonne National Laboratory (ANL) to compute the core-wide temperature profiles in Liquid Metal Reactor (LMR) cores. The accuracy of this code has recently been tested by comparing the predicted temperatures with measured values in the Experimental Breeder Reactor R (EBR-II). The detailed temperature distributions in two experimental subassemblies and the mixed mean subassembly outlet temperatures were used in this validation study. The SE2-ANL predictions were found to agree well with measured values. It was also found that SE2-ANL yields results with accuracy comparable to the more detailed COBRA-WC [2] calculations at much lower computational cost.
Date: January 1995
Creator: Yang, W. S. & Yacout, A. M.
Partner: UNT Libraries Government Documents Department

Effects of Pb and Bi cross sections on ATW subcriticality predictions.

Description: The accelerator-driven transmutation of waste (ATW) system has been proposed for transmuting the long-lived radioactive nuclei of high-level waste to stable or short-lived species. In recent ATW design concepts, lead-bismuth eutectic (LBE), consisting of 44.5% Pb and 55.5% Bi by weight is used as the spallation target, system coolant, and reflector. Because of the excellent neutron reflection properties of LBE, the subcriticality level of ATW is quite sensitive to the cross sections of lead and bismuth. The purpose of this paper is to investigate the effects of these cross sections on subcriticality and other core characteristics of ATW and to compare the results obtained using cross sections in different evaluated nuclear data files. The effects of lead and bismuth cross sections on the core characteristics of ATW were studied using 33 group cross section sets derived from the ENDF/B-VI, ENDF/B-V, JENDL-3.2, and BROND-2.2 nuclear data. A 2000 MW(thermal) ATW configuration similar to that described in Reference 1 was used in this study. In this configuration, the spallation target region is 55 cm high and 25 cm in radius, and is surrounded by a 15-cm thick LBE buffer. The adjacent fueled region is {approximately}65 cm thick and 200 cm high. The volume fractions of fuel, coolant, and structure are 25.7%, 59.3%, and 15%, respectively. The metal alloy fuel is composed of roughly 70% zirconium, 25% transuranics (TRU), and 5% Tc-99 by weight. A thick LBE reflector surrounds the whole core; its axial thickness is 250 cm, and its radial thickness is 295.2 cm.
Date: June 25, 1999
Creator: Khalil, H. S. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

Analysis of the ATW fuel cycle using the REBUS-3 code system.

Description: Partitioning and transmutation strategies are under study in several countries as a means of reducing the long-term hazards of spent fuel and other high-level nuclear waste. Various reactor and accelerator-driven system concepts have been proposed to transmute the long-lived radioactive nuclei of waste into stable or short-lived species. Among these concepts, the accelerator-driven transmutation of waste (ATW) system has been proposed by LANL for rapid destruction of transuranic actinides and long-lived fission products ({sup 99}Tc and {sup 129}I).The current reference ATW concept employs a subcritical, liquid metal cooled, fast-spectrum nuclear subsystem. Because the discharged fuel is recycled, analysis of ATW nuclear performance requires modeling of the external cycle as well as the in-core fuel management. The fuel cycle analysis of ATW can be performed rigorously using Monte Carlo calculations coupled with detailed depletion calculations. However, the inefficiency of this approach makes it impractical, particularly in view of (a) the large number of fuel cycle calculations needed for design optimization and (b) the need to represent complex in-core and out-of-core fuel cycle operations. To meet the need for design-oriented capabilities, tools previously developed for fast reactor calculations are being adapted for application to ATW. Here we describe the extension and application of the REBUS-3 code to ATW fuel cycle analysis. This code has been extensively used for advanced liquid metal reactor design and analysis and validated against EBR-II irradiation data.
Date: June 25, 1999
Creator: Khalil, H. S. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

Preliminary neutronic studies for the liquid-salt-cooled very hightemperature reactor (LS-VHTR).

Description: Preliminary neutronic studies have been performed in order to provide guidelines to the design of a liquid-salt cooled Very High Temperature Reactor (LS-VHTR) using Li{sub 2}BeF{sub 4} (FLiBe) as coolant and a solid cylindrical core. The studies were done using the lattice codes (WIMS8 and DRAGON) and the linear reactivity model to estimate the core reactivity balance, fuel composition, discharge burnup, and reactivity coefficients. An evaluation of the lattice codes revealed that they give very similar accuracy as the Monte Carlo MCNP4C code for the prediction of the fuel element multiplication factor (kinf) and the double heterogeneity effect of the coated fuel particles in the graphite matrix. The loss of coolant from the LS-VHTR core following coolant voiding was found to result in a positive reactivity addition, due primarily to the removal of the strong neutron absorber Li-6. To mitigate this positive reactivity addition and its impact on reactor design (positive void reactivity coefficient), the lithium in the coolant must be enriched to greater than 99.995% in its Li-7 content. For the reference LS-VHTR considered in this work, it was found that the magnitude of the coolant void reactivity coefficient (CVRC) is quite small (less than $1 for 100% voiding). The coefficient was found to become more negative or less positive with increase in the lithium enrichment (Li-7 content). It was also observed that the coefficient is positive at the beginning of cycle and becomes more negative with increasing burnup, indicating that by using more than one fuel batch, the coefficient could be made negative at the beginning of cycle. It might, however, still be necessary at the beginning of life to design for a negative CVRC value. The study shows that this can be done by using burnable poisons (erbium is a leading candidate) or by changing the reference ...
Date: October 5, 2005
Creator: Kim, T. K.; Taiwo, T. A. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

Description: Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC{sup 2}-2/TWODANT results ...
Date: May 16, 2008
Creator: Yang, W. S. & Lee, C. H. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Interim report on fuel cycle neutronics code development.

Description: As part of the Global Nuclear Energy Partnership (GNEP), a fast reactor simulation program was launched in April 2007 to develop a suite of modern simulation tools specifically for the analysis and design of sodium cooled fast reactors. The general goal of the new suite of codes is to reduce the uncertainties and biases in the various areas of reactor design activities by enhanced prediction capabilities. Under this fast reactor simulation program, a high-fidelity deterministic neutron transport code named UNIC is being developed. The final objective is to produce an integrated, advanced neutronics code that allows the high fidelity description of a nuclear reactor and simplifies the multi-step design process by direct coupling with thermal-hydraulics and structural mechanics calculations. Currently there are three solvers for the neutron transport code incorporated in UNIC: PN2ND, SN2ND, and MOCFE. PN2ND is based on a second-order even-parity spherical harmonics discretization of the transport equation and its primary target area of use is the existing homogenization approaches that are prevalent in reactor physics. MOCFE is based upon the method of characteristics applied to an unstructured finite element mesh and its primary target area of use is the fine grained nature of the explicit geometrical problems which is the long term goal of this project. SN2ND is based on a second-order, even-parity discrete ordinates discretization of the transport equation and its primary target area is the modeling transition region between the PN2ND and MOCFE solvers. The major development goal in fiscal year 2008 for the MOCFE solver was to include a two-dimensional capability that is scalable to hundreds of processors. The short term goal of this solver is to solve two-dimensional representations of reactor systems such that the energy and spatial self-shielding are accounted for and reliable cross sections can be generated for the homogeneous calculations. ...
Date: May 13, 2008
Creator: Rabiti, C; Smith, M. A.; Kaushik, D. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

A validation study of existing neutronics tools against ZPPR-21 and ZPPR-15 critical experiments.

Description: A study was performed to validate the existing tools for fast reactor neutronics analysis against previous critical experiments. The six benchmark problems for the ZPPR-21 critical experiments phases A through F specified in the Handbook of Evaluated Criticality Safety Benchmark Experiments were analyzed. Analysis was also performed for three loading configurations of the ZPPR-15 Phase A experiments. As-built core models were developed in XYZ geometries using the reactor loading records and drawer master information. Detailed Monte Carlo and deterministic transport calculations were performed, along with various modeling sensitivity analyses. The Monte Carlo simulations were carried out with the VIM code with continuous energy cross sections based on the ENDF/B-V.2 data. For deterministic calculations, region-dependent 230-group cross sections were generated using the ETOE-2/MC-2/SDX code system, again based on the ENDF/B-V.2 data. Plate heterogeneity effects were taken into account by SDX unit cell calculations. Core calculations were performed with the TWODANT discrete ordinate code for the ZPPR-21 benchmarks, and with the DIF3D nodal transport option for the ZPPR-15 experiments. For all six ZPPR-21 configurations where the Pu-239 concentration varies from 0 to 49 w/o and the U-235 concentration accordingly varies from 62 to 0 w/o, the core multiplication factor determined with a 230-group TWODANT calculation agreed with the VIM Monte Carlo solution within 0.20 %{Delta}k, and there was no indication of any systematic bias. The quality of principal cross sections generated with the MC-2 code was comparable to that of VIM cross sections. The overall reactivity effect due to the errors in the 230-group principal cross sections was estimated to be less than 0.05 %{Delta}k. The statistics of the differences between calculated values and specified benchmark experimental values showed similar bias (from -0.28 %{Delta}k to 0.33 %{Delta}k) for MC{sup 2}-2/TWODANT and VIM. This result suggests that the criticality prediction accuracy of MC{sup ...
Date: September 30, 2007
Creator: Yang, W.S. & Kim, S.J. (Nuclear Engineering Division)
Partner: UNT Libraries Government Documents Department

Reactivity estimation for source-driven systems using first-order perturbation theory.

Description: Applicability of the first-order perturbation (FOP) theory method to reactivity estimation for source-driven systems is examined in this paper. First, the formally exact point kinetics equations have been derived from the space-dependent kinetics equations and the kinetics parameters including the dynamic reactivity have been defined. For the dynamic reactivity, exact and first-order perturbation theory expressions for the reactivity change have been formulated for source-driven systems. It has been also shown that the external source perturbation itself does not change the reactivity if the initial {lambda}-mode adjoint flux is used as the weight function. Using two source-driven benchmark problems, the reactivity change has been estimated with the FOP theory method for various perturbations. By comparing the resulting reactivity changes with the exact dynamic reactivity changes determined from the space-dependent kinetics solutions, it has been shown that the accuracy of the FOP theory method for the accelerator-driven system (ADS) is reasonably good and comparable to that for the critical reactors. The adiabatic assumption has also been shown to be a good approximation for the ADS kinetics analyses.
Date: July 2, 2002
Creator: Kim, Y.; Yang, W. S.; Taiwo, T. A. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

ATW neutronics design studies.

Description: The Accelerator Transmutation of Waste (ATW) concept has been proposed as a transuranics (TRU) (and long-lived fission product) incinerator for processing the 87,000 metric tonnes of Light Water Reactor used fuel which will have been generated by the time the currently deployed fleet of commercial reactors in the US reach the end of their licensed lifetime. The ATW is proposed to separate the uranium from the transuranics and fission products in the LWR used fuel, to fission the transuranics, to send the LWR and ATW generated fission products to the geologic repository and to send the uranium to either a low level waste disposal site or to save it for future use. The heat liberated in fissioning the transuranics would be converted to electricity and sold to partially offset the cost of ATW construction and operations. Options for incineration of long-lived fission products are under evaluation. A six-year science-based program of ATW trade and system studies was initiated in the US FY 2000 to achieve two main purposes: (1) ''to evaluate ATW within the framework of nonproliferation, waste management, and economic considerations,'' and (2) ''to evaluate the efficacy of the numerous technical options for ATW system configuration.'' This paper summarizes the results from neutronics and thermal/hydraulics trade studies which were completed at Argonne National Laboratory during the first year of the program. Core designs were developed for Pb-Bi cooled and Na cooled 840 MW{sub th} fast spectrum transmuter designs employing recycle. Additionally, neutronics analyses were performed at Argonne for a He cooled 600 MW{sub th} hybrid thermal and fast core design proposed by General Atomics Co. which runs critical for 3/4 and subcritical for 1/4 of its four year once-thin burn cycle. The mass flows and the ultimate loss of transuranic isotopes to the waste stream per unit of heat ...
Date: November 10, 2000
Creator: Wade, D. C.; Yang, W. S. & Khalil, H.
Partner: UNT Libraries Government Documents Department

The impact of covariance information on criticality safety calculations in the resolved resonance energy range.

Description: Resonance data play a significant role in the calculations of systems considered for criticality safety applications. K{sub eff}, the major parameter of interest in such a type of calculations, can be heavily dependent both on the quality of the resonance data as well as on the accuracy achieved in the processing of these data. If reasonable uncertainty values are available, in conjunction with their correlation in energy and among type of resonance parameters, one can exploit existing methodologies, based on perturbation theory, in order to evaluate their impact on the integral parameter of interest, i.e., K{sub eff} in our case, in practical applications. In this way, one could be able to judge if the uncertainty on specific quantities, e.g., covariances on resonance data, have a significant impact and, therefore, deserve a careful evaluation. This report, first, will recall the basic principles that lie behind an uncertainty evaluation and review the current situation in the field of covariance data. Then an attempt is made for defining a methodology that allows calculating covariances values for resolved resonance parameters. Finally, practical applications, of interest for criticality safety calculations, illustrate the impact of different assumptions on correlations among resolved resonance parameters.
Date: June 11, 2004
Creator: Naberejnev, D. G.; Palmiotti, G. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

A feasibility study of reactor-based deep-burn concepts.

Description: A systematic assessment of the General Atomics (GA) proposed Deep-Burn concept based on the Modular Helium-Cooled Reactor design (DB-MHR) has been performed. Preliminary benchmarking of deterministic physics codes was done by comparing code results to those from MONTEBURNS (MCNP-ORIGEN) calculations. Detailed fuel cycle analyses were performed in order to provide an independent evaluation of the physics and transmutation performance of the one-pass and two-pass concepts. Key performance parameters such as transuranic consumption, reactor performance, and spent fuel characteristics were analyzed. This effort has been undertaken in close collaborations with the General Atomics design team and Brookhaven National Laboratory evaluation team. The study was performed primarily for a 600 MWt reference DB-MHR design having a power density of 4.7 MW/m{sup 3}. Based on parametric and sensitivity study, it was determined that the maximum burnup (TRU consumption) can be obtained using optimum values of 200 {micro}m and 20% for the fuel kernel diameter and fuel packing fraction, respectively. These values were retained for most of the one-pass and two-pass design calculations; variation to the packing fraction was necessary for the second stage of the two-pass concept. Using a four-batch fuel management scheme for the one-pass DB-MHR core, it was possible to obtain a TRU consumption of 58% and a cycle length of 286 EFPD. By increasing the core power to 800 MWt and the power density to 6.2 MW/m{sup 3}, it was possible to increase the TRU consumption to 60%, although the cycle length decreased by {approx}64 days. The higher TRU consumption (burnup) is due to the reduction of the in-core decay of fissile Pu-241 to Am-241 relative to fission, arising from the higher power density (specific power), which made the fuel more reactivity over time. It was also found that the TRU consumption can be improved by utilizing axial fuel shuffling ...
Date: September 16, 2005
Creator: Kim, T. K.; Taiwo, T. A.; Hill, R. N. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

A particle-bed gas cooled fast reactor core design for waste minimization.

Description: The issue of waste minimization in advanced reactor systems has been investigated using the Particle-Bed Gas-Cooled Fast Reactor (PB-GCFR) design being developed and funded under the U.S. Department of Energy Nuclear Energy Research Initiative (USDOE NERI) Program. Results indicate that for the given core power density and constraint on the maximum TRU enrichment allowable, the lowest amount of radiotoxic transuranics to be processed and hence sent to the repository is obtained for long-life core designs. Calculations were additionally done to investigate long-life core designs using LWR spent fuel TRU and recycle TRU, and different feed, matrix and reflector materials. The recycled TRU and LWR spent TRU fuels give similar core behaviors, because of the fast spectrum environment which does not significantly degrade the TRU composition. Using light elements as reflector material was found to be unattractive because of power peaking problems and large reactivity swings. The application of a lead reflector gave the longest cycle length and lowest TRU processing requirement. Materials compatibility and performance issues require additional investigation.
Date: October 11, 2002
Creator: Hoffman, E. A.; Taiwo, T. A.; Yang, W. S. & Fatone, M.
Partner: UNT Libraries Government Documents Department

Effects of buffer thickness on ATW blanket performance.

Description: This paper presents preliminary results of target and buffer design studies for liquid metal cooled accelerator transmutation of waste (ATW) systems, aimed at maximizing the source importance while simultaneously reducing the irradiation damage to fuel. Using 840 MWt liquid metal cooled ATW designs, the effects of buffer thickness on the blanket performance have been studied. Varying the buffer thickness for a given blanket configuration, system performance parameters have been estimated by a series of calculations using the MCNPX and REBUS-3 codes. The effects of source importance variation are studied by investigating the low-energy (< 20 MeV) neutron source distribution and the equilibrium cycle blanket performance parameters such as fuel inventory, discharge burnup, burnup reactivity loss, and peak fast fluence. For investigating irradiation damage to fuel, the displacements per atom (dpa), hydrogen production, and helium production rates are evaluated at the buffer and blanket interface where the peak fast fluence occurs. Results for the liquid-metal-cooled designs show that the damage rates and the source importance increase monotonically as the buffer thickness decreases. Based on a compromise between the competing objectives of increasing the source importance and reducing the damage rates, a buffer thickness of around 20 cm appears to be reasonable. Investigation of the impact of the proton beam energy on the target and buffer design shows that for a given blanket power level, a lower beam energy (0.6 GeV versus 1 GeV) results in a higher irradiation damage to the beam window. This trend occurs because of the increase in the beam intensity required to maintain the power level.
Date: August 10, 2001
Creator: Yang, W. S.; Mercatali, L.; Taiwo, T. A. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

Direct solution of the mathematical adjoint equations for an interface current nodal formulation

Description: A numerical method for directly computing the mathematical adjoint flux moments and partial currents for the hexagonal-Z geometry interface current nodal formulation in the DIF3D code is described. The new scheme is developed as an alternative to an existing scheme that employs a similarity transformation of the physical adjoint solution to compute the mathematical adjoint. Whereas the existing scheme is rigorous only when the flat transverse-leakage approximation is employed, this new scheme is exact for all leakage approximations in the DIF3D nodal method. in the new scheme, adjoint nodal equations whose form is very similar to that of the forward nodal equations are derived by employing linear combinations of the adjoint partial currents as computational unknowns in the adjoint equations. This enables the use of the forward solution algorithm with only minor modifications for solving the mathematical adjoint equations. By using the new scheme as a reference method, it is shown numerically that while the results computed with the existing scheme are approximate, they are sufficiently accurate for calculations of global and local reactivity changes resulting from coolant voiding in a liquid metal reactor.
Date: January 1, 1993
Creator: Taiwo, T.A.; Yang, W.S. & Khalil, H.S.
Partner: UNT Libraries Government Documents Department

Preliminary estimation of isotopic inventories of 2000 MWt ABR (revision 1).

Description: The isotopic inventories of a 2000 MWt Advanced Burner Reactor (ABR) core have been estimated to support the ABR accident analysis to be reported in the Appendix D of the Programmatic Environmental Impact Statement (PEIS). Based on the Super-PRISM design, a preliminary core design of 2000 MWt ABR was developed to achieve a one-year cycle length with 3-batch fuel management scheme. For a bounding estimation of transuranics (TRU) inventory, a low TRU conversion ratio ({approx}0.3) was targeted to increase the TRU enrichment. By changing the fuel compositions, isotopic inventories of mass and radioactivity were evaluated for four different core configurations: recycled metal fuel core, recycled oxide fuel core, startup metal fuel core, and startup oxide fuel core. For recycled cores, the TRU recovered from ABR spent fuel was used as the primary TRU feed, and the TRU recovered from 10-year cooled light water reactor spent fuel was used as the makeup TRU feed. For startup cores, weapons-grade plutonium was used as TRU feed without recycling ABR spent fuel. It was also assumed that a whole batch of discharged fuel assemblies is stored in the in-vessel storage for an entire irradiation cycle. For both metal and oxide fuel cores, the estimated TRU mass at beginning of equilibrium cycle (BOEC), including spent fuel TRU stored in the in-vessel storage, was about 8.5-8.7 MT for the recycled cores and 5.2 MT for the startup cores. Since a similar power was generated, the fission product mass are comparable for all four cores: 1.4 MT at BOEC and about 2.0 MT at end of equilibrium cycle (EOEC). Total radioactivity at BOEC is about 8.2 x 10{sup 8} curies in recycled cores and about 6.9 x 10{sup 8} curies in startup cores, and increases to about 1.1 x 10{sup 10} curies at EOEC for all four ...
Date: June 16, 2008
Creator: Kim, T. K. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

Development of a coupled dynamics code with transport theory capability and application to accelerator driven systems transients

Description: The VARIANT-K and DIF3D-K nodal spatial kinetics computer codes have been coupled to the SAS4A and SASSYS-1 liquid metal reactor accident and systems analysis codes. SAS4A and SASSYS-1 have been extended with the addition of heavy liquid metal (Pb and Pb-Bi) thermophysical properties, heat transfer correlations, and fluid dynamics correlations. The coupling methodology and heavy liquid metal modeling additions are described. The new computer code suite has been applied to analysis of neutron source and thermal-hydraulics transients in a model of an accelerator-driven minor actinide burner design proposed in an OECD/NEA/NSC benchmark specification. Modeling assumptions and input data generation procedures are described. Results of transient analyses are reported, with emphasis on comparison of P1 and P3 variational nodal transport theory results with nodal diffusion theory results, and on significance of spatial kinetics effects.
Date: March 9, 2000
Creator: Cahalan, J. E.; Ama, T.; Palmiotti, G.; Taiwo, T. A. & Yang, W. S.
Partner: UNT Libraries Government Documents Department

An investigation of an optimal range of subcriticality for accelerator - driven systems.

Description: It is attempted in this paper to define an optimal range of subcriticality of ADS systems from the operational and safety points of view. To devise a representative measure of the subcriticality level, the mathematical and physical implications of the effective multiplication factor and the source multiplication factor have been reviewed. A set of criteria that bound the feasible subcriticality level is proposed in terms of the effective multiplication factor; the minimum required subcriticality is determined by the largest value of potential reactivity increase including the temperature defect and the calculation and measurement uncertainties, and the maximum allowable subcriticality level is bounded by the system economy and the technical feasibility of the system. Within this feasible domain of subcriticality, a preliminary estimation of the optimal range of subcriticality was performed for a lead-bismuth-eutectic (LBE) cooled ADS design based on the safety and transmutation performances. The effects on the system safety of the subcriticality level were analyzed for several important transients using an integral safety analysis method, and the transmutation performance was evaluated in terms of the fuel and long-lived fission product discharge burnups.
Date: July 2, 2002
Creator: Kim, Y.; Park, W. S.; Yang, W. S.; Taiwo, T. A. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

A Global Approach to the Physics Validation of Simulation Codes for Future Nuclear Systems

Description: This paper presents a global approach to the validation of the parameters that enter into the neutronics simulation tools for advanced fast reactors with the objective to reduce the uncertainties associated to crucial design parameters. This global approach makes use of sensitivity/uncertainty methods; statistical data adjustments; integral experiment selection, analysis and “representativity” quantification with respect to a reference system; scientifically based cross section covariance data and appropriate methods for their use in multigroup calculations. This global approach has been applied to the uncertainty reduction on the criticality of the Advanced Burner Reactor, (both metal and oxide core versions) presently investigated in the frame of the GNEP initiative. The results obtained are very encouraging and allow to indicate some possible improvements of the ENDF/B-VII data file.
Date: September 1, 2008
Creator: Palmiotti, Giuseppe; Salvatores, Massimo; Aliberti, Gerardo; Hiruta, Hikarui; McKnight, R.; Oblozinsky, P. et al.
Partner: UNT Libraries Government Documents Department