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The steady-state thermal-hydraulic performance of 3500 MWth metal and oxide fueled LMRs

Description: The thermal-hydraulic performance of a 3500 MWth metal and oxide fueled LMR is reported. Orifice zones are defined and coolant flowrates are given for use in safety analyses. The flux calculations were carried out in three-dimensional hexagonal-Z geometry using a finite differenced diffusion theory code. The heating calculations included the transport and deposition of gamma energy. The assembly temperature calculations were performed using a subchannel code.
Date: March 1, 1989
Creator: Vilim, R.B. & Hill, R.N.
Partner: UNT Libraries Government Documents Department

LMR design concepts for transuranic management in low sodium void worth cores

Description: The fuel cycle processing techniques and hard neuron spectrum of the Integral Fast Reactor (IFR) metal fuel cycle have favorable characteristics for the management of transuranics; and the wide range of breeding characteristics available in metal fuelled cores provides for flexibility in transuranic management strategy. Previous studies indicate that most design options which decrease the breeding ratio also show a decrease in sodium void worth; therefore, low void worths are achievable in transuranic burning (low breeding ratio) core designs. This paper describes numerous trade studies assessing various design options for a low void worth transuranic burner core. A flat annular core design appears to be a promising concept; the high leakage geometry yields a low breeding ratio and small sodium void worth. To allow flexibility in breeding characteristics, alternate design options which achieve fissile self-sufficiency are also evaluated. A self-sufficient core design which is interchangeable with the burner core and maintains a low sodium void worth is developed. 13 refs., 1 fig., 4 tabs.
Date: January 1, 1991
Creator: Hill, R.N.
Partner: UNT Libraries Government Documents Department

Advanced methods comparisons of reaction rates in the Purdue Fast Breeder Blanket Facility

Description: A review of worldwide results revealed that reaction rates in the blanket region are generally underpredicted with the discrepancy increasing with penetration; however, these results vary widely. Experiments in the large uniform Purdue Fast Breeder Blanket Facility (FBBF) blanket yield an accurate quantification of this discrepancy. Using standard production code methods (diffusion theory with 50 group cross sections), a consistent Calculated/Experimental (C/E) drop-off was observed for various reaction rates. A 50% increase in the calculated results at the outer edge of the blanket is necessary for agreement with experiments. The usefulness of refined group constant generation utilizing specialized weighting spectra and transport theory methods in correcting this discrepancy was analyzed. Refined group constants reduce the discrepancy to half that observed using the standard method. The surprising result was that transport methods had no effect on the blanket deviations; thus, transport theory considerations do not constitute or even contribute to an explanation of the blanket discrepancies. The residual blanket C/E drop-off (about half the standard drop-off) using advanced methods must be caused by some approximations which are applied in all current methods. 27 refs., 3 figs., 1 tab.
Date: January 1, 1988
Creator: Hill, R.N. & Ott, K.O.
Partner: UNT Libraries Government Documents Department

Performance assessment modeling of radionuclide transport: Matrix/Fracture and colloidal transport

Description: Several total system PA analyses have been completed to investigate the performance of light water reactor (LWR) SNF or alternative waste forms in a geologic repository. These analyses contained either no modeling or simple modeling of fracture flow and transport; and none considered radio-colloid facilitated transport. This paper summarizes the work completed to develop a transport model which considers fracture, matrix, and radio-colloid transport; this model is used to evaluate the transport of SNF radionuclides at the Yucca Mountain site.
Date: April 1, 1996
Creator: Nutt, W.M. & Hill, R.N.
Partner: UNT Libraries Government Documents Department

Summary of Generation-IV transmutation impacts.

Description: An assessment of the potential role of Generation IV nuclear systems in an advanced fuel cycle has been performed. The Generation IV systems considered are the thermal-spectrum VHTR and SCWR, and the fast-spectrum GFR, LFR, and SFR. This report addresses the impact of each system on advanced fuel cycle goals, particularly related to waste management and resource utilization. The transmutation impact of each system was also assessed, along with variant designs for transuranics (TRU) burning. The base fuel cycle for the thermal reactor concepts (VHTR and SCWR) is a once-through fuel cycle using low-enriched uranium fuels. The higher burnup and thermal efficiency of the VHTR gives an advantage in terms of heavy-metal waste mass and volume, with lower decay heat and radiotoxicity of the spent fuel per electrical energy produced, compared to a PWR. Fuel utilization might, however, be worse compared to the PWR, because of the higher fuel enrichment essential to meeting the VHTR system design requirements. The SCWR concept also featured improved thermal efficiency; however, benefits are reduced by the lower fuel discharge burnup. The base fuel cycle for the fast reactor concepts (SFR, GFR, and LFR) is a closed fuel cycle using recycled TRU and depleted uranium fuels. Waste management gains from complete recycle are substantial, with the final disposition heat load determined by processing losses. The base Generation-IV concepts allow consumption of U-238 significantly extending uranium resources (up to 100 times). For both thermal and fast concepts, recent design studies have pursued the development of dedicated burner designs. Preliminary results suggest that a burnup of 50-60% is possible in a VHTR burner design using non-uranium (transuranics) fuel. However, practical limits related to higher actinide buildup and safety impact may limit the extent of TRU burning in thermal reactors. Fast burner designs have been developed for both ...
Date: August 3, 2005
Creator: Taiwo, T. A. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

Integral transport treatment of transitional resonance spectra

Description: Analytical exploratory investigations indicated that transition effects such as streaming will cause a considerable spatial variation in the neutron spectra across resonances; streaming leads to opposite effects in the forward and backward directions. The neglect of this spatial and angular variation of the transitory resonance spectra is an approximation that is common to all current methodologies. An integral transport theory formalism was developed for the description of spatially dependent spectra in isolated resonances. This treatment differentiates between forward and backward directed components of the neutron flux in slab geometry. This theory was applied to an isolated actinide resonance in a simplified fast reactor blanket problem. The resonance spectra of the directional flux components, /phi//sup +/ and /phi//sup -/, and even more so the 90/degree/ cone components, were shown to deviate significantly from the infinite medium approximation with the differences increasing with penetration. The changes in /phi//sup +/ lead to a decreasing scattering group constant which enhances neutron transmission; the changes in /phi//sup -/ lead to an increasing group constant inhibiting backward scattering. Therefore, the changes in the forward and backward directed spectra both lead to increased neutron transmission. Conversely, the flux (/phi/=/phi//sup +/+/phi//sup -/) was shown both in the analytical formulas and in the numerical solution to agree closely with the infinite medium approximation; the directional effects cancel in the summation. Therefore, flux-weighted (''diffusion theory'') group constants cannot yield the required increase in transmission even using transport theory. 8 refs., 2 figs.
Date: January 1, 1988
Creator: Hill, R.N.; Ott, K.O. & Rhodes, J.D.
Partner: UNT Libraries Government Documents Department

IAEA sodium void reactivity benchmark calculations

Description: In this paper, the IAEA-1 992 Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core'' problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated.
Date: January 1, 1992
Creator: Hill, R.N. & Finck, P.J.
Partner: UNT Libraries Government Documents Department

Development and analysis of a compact low-conversion ratio fast burner reactor.

Description: This report explores design options for compact fast burner reactors that can achieve low conversion ratios. Operational characteristics and whole-core reactivity coefficients are generated and contrasted with low conversion ratio designs of previous studies. A compact core point design is then selected and detailed reactivity coefficients are displayed and discussed. The effectiveness of fast spectrum systems for actinide transmutation has been well documented. The key advantage of the fast spectrum resides in the severely reduced capture/fission ratios. this inhibits the production of the higher actinides that dominate the long-term radiotoxicity of nuclear waste. In conventional fast burner studies, the transmutation rate was limited by constraints placed on the fuel composition. In an earlier phase of this study the entire range of fuel compositions (including non-uranium fuel) was explored to assess the performance and safety limits of fast burner reactor systems. In this report, similar fuel compositions are utilized for application in compact configurations to achieve conversion ratios below 0.5.
Date: May 12, 2006
Creator: Smith, M. A. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

Development of small, fast reactor core designs using lead-based coolant.

Description: A variety of small (100 MWe) fast reactor core designs are developed, these include compact configurations, long-lived (15-year fuel lifetime) cores, and derated, natural circulation designs. Trade studies are described which identify key core design issues for lead-based coolant systems. Performance parameters and reactivity feedback coefficients are compared for lead-bismuth eutectic (LBE) and sodium-cooled cores of consistent design. The results of these studies indicate that the superior neutron reflection capability of lead alloys reduces the enrichment and burnup swing compared to conventional sodium-cooled systems; however, the discharge fluence is significantly increased. The size requirement for long-lived systems is constrained by reactivity loss considerations, not fuel burnup or fluence limits. The derated lead-alloy cooled natural circulation cores require a core volume roughly eight times greater than conventional compact systems. In general, reactivity coefficients important for passive safety performance are less favorable for the larger, derated configurations.
Date: June 11, 1999
Creator: Cahalan, J. E.; Hill, R. N.; Khalil, H. S. & Wade, D. C.
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

An evaluation of waste radiotoxicity reduction for a fast burner reactor closed fuel cycle: NEA benchmark results

Description: As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed. In this paper, the fuel cycle performance of the metal-fueled benchmark is evaluated in detail. Benchmark results assess the reactor performance and toxicity behavior in a closed nuclear fuel cycle for a parametric variation of the conversion ratio between 0.5 and 1.0. Results indicate that a fast burner reactor closed fuel cycle can be utilized to significantly reduce the radiotoxicity destined for ultimate disposal.
Date: December 1, 1995
Creator: Grimm, K.N.; Hill, R.N. & Wase, D.C.
Partner: UNT Libraries Government Documents Department

Fast burner reactor benchmark results from the NEA working party on physics of plutonium recycle

Description: As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed; fuel cycle scenarios using either PUREX/TRUEX (oxide fuel) or pyrometallurgical (metal fuel) separation technologies were specified. These benchmarks were designed to evaluate the nuclear performance and radiotoxicity impact of a transuranic-burning fast reactor system. International benchmark results are summarized in this paper; and key conclusions are highlighted.
Date: December 1, 1995
Creator: Hill, R.N.; Wade, D.C. & Palmiotti, G.
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

Assessment of transuranics stabilization in PWRs

Description: The stabilization of transuranics (TRU) in a PWR fuel cycle was evaluated for the CORAIL assembly. Alternative assembly designs (a highly moderated and modified CORAIL-TRU assembly and a homogeneous Thorium-TRU assembly) were also investigated to assess the potential of obtaining a near-zero TRU mass balance (i.e., the net TRU production per assembly) and low power peaking factor. The radiotoxicity of the nuclear waste sent to the repository environment and the impact of TRU stabilization on the future TRU stockpile were also evaluated. Assembly level mass flow analyses have shown that TRU mass balances in the range of 0.2 to 1.4 kg/assembly are achievable within 7 recycles of the TRU, compared with 6.5 kg/assembly for a reference UO{sub 2} assembly. The study also revealed that the radiotoxicity of the repository waste generated by these TRU-containing assemblies at 10 years after disposal is roughly half that of a reference UO{sub 2} assembly; furthermore, the radiotoxicity falls below that of natural uranium ore after about 500 years because only a small fraction of the TRU (0.1%) is passed to the waste repository. Finally, the future TRU stockpile could be reduced by implementation of TRU multi-recycling in the CORAIL or alternative assemblies in a current-generation PWR core.
Date: July 16, 2002
Creator: Kim, T. K.; Stillman, J. A.; Taiwo, T. A. & Hill, R. N.
Partner: UNT Libraries Government Documents Department

Performance assessment modeling of high level nuclear wasteforms from the pyroprocess fuel cycle

Description: Several performance assessment (PA) analyses have been completed to estimate the release to the accessible environment of radionuclides from spent light water reactor (LWR) fuel emplaced in the proposed Yucca Mountain repository. Probabilistic methods were utilized based on the complexity of the repository system. Recent investigations have been conducted to identify the merits of a pyroprocess fuel cycle. This cycle utilizes high temperature molten salts and metals to partially separate actinides and fission products. In a closed liquid metal reactor (LMR) fuel cycle, this allows recycling of nearly all of the actinides. In a once-through cycle, this isolates the actinides for storage into a wasteform which can be specifically tailored for their retention. With appropriate front-end treatment, this Process can also be used to treat LWR spent fuel.
Date: June 1, 1995
Creator: Nutt, W.M.; Hill, R.N. & Bullen, D.B.
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

A comparison of equilibrium and non-equilibrium cycle methods for Na-cooled ATW system.

Description: An equilibrium cycle method, embodied in the REBUS-3[1] code system, has generally been used in conventional fast reactor design activities. The equilibrium cycle method provides an efficient approach for modeling reactor system, compared to the more traditional non-equilibrium cycle fuel management calculation approach. Recently, the equilibrium analysis method has been utilized for designing Accelerator Transmutation of Waste (ATW)[2,3,4] cores, in which a scattered-reloading fuel management scheme is used. Compared with the conventional fast reactors, the ATW core is significantly different in several aspects since its main mission is to incinerate the transuranic (TRU) fuels. The high burnup non-fertile fuel has large variations in composition and reactivity during its lifetime. Furthermore, a relatively short cycle length is utilized in the ATW design to limit the potentially large reactivity swing over a cycle, and consequently 7 or 8-batch fuel management is usually assumed for a high fuel burnup. The validity of the equilibrium analysis method for the ATW core, therefore, needed to be verified. The main objective of this paper is to assess the validity of the equilibrium analysis method for a Na-cooled ATW core[4], which is an alternative core design of the ATW system under development.
Date: March 30, 2002
Creator: Kim, Y.; Hill, R. N. & Taiwo, T. A.
Partner: UNT Libraries Government Documents Department

Physics studies of higher actinide consumption in an LMR

Description: The core physics aspects of the transuranic burning potential of the Integral Fast Reactor (IFR) are assessed. The actinide behavior in fissile self-sufficient IFR closed cycles of 1200 MWt size is characterized, and the transuranic isotopics and risk potential of the working inventory are compared to those from a once-through LWR. The core neutronic performance effects of rare-earth impurities present in the recycled fuel are addressed. Fuel cycle strategies for burning transuranics from an external source are discussed, and specialized actinide burner designs are described. 4 refs., 4 figs., 3 tabs.
Date: January 1, 1990
Creator: Hill, R.N.; Wade, D.C.; Fujita, E.K. & Khalil, H.S.
Partner: UNT Libraries Government Documents Department

An evaluation of multigroup flux predictions in the EBR-II core

Description: The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.
Date: January 1, 1991
Creator: Hill, R.N.; Fanning, T.H. & Finck, P.J.
Partner: UNT Libraries Government Documents Department

Improvements in EBR-2 core depletion calculations

Description: The need for accurate core depletion calculations in Experimental Breeder Reactor No. 2 (EBR-2) is discussed. Because of the unique physics characteristics of EBR-2, it is difficult to obtain accurate and computationally efficient multigroup flux predictions. This paper describes the effect of various conventional and higher order schemes for group constant generation and for flux computations; results indicate that higher-order methods are required, particularly in the outer regions (i.e. the radial blanket). A methodology based on Nodal Equivalence Theory (N.E.T.) is developed which allows retention of the accuracy of a higher order solution with the computational efficiency of a few group nodal diffusion solution. The application of this methodology to three-dimensional EBR-2 flux predictions is demonstrated; this improved methodology allows accurate core depletion calculations at reasonable cost. 13 refs., 4 figs., 3 tabs.
Date: January 1, 1991
Creator: Finck, P.J.; Hill, R.N. & Sakamoto, S.
Partner: UNT Libraries Government Documents Department

An evaluation of multigroup flux predictions in the EBR-II core

Description: The unique physics characteristics of EBR-II which are difficult to model with conventional neutronic methodologies are identified; the high neutron leakage fraction and importance of neutron reflection cause errors when conventional calculational approximations are utilized. In this paper, various conventional and higher-order group constant evaluations and flux computation methods are compared for a simplified R-Z model of the EBR-II system. Although conventional methods do provide adequate predictions of the flux in the core region, significant mispredictions are observed in the reflector and radial blanket regions. Calculational comparisons indicate that a fine energy group structure is required for accurate predictions of the eigenvalue and flux distribution; greater detail is needed in the iron resonance scattering treatment. Calculational comparisons also indicate that transport theory with detailed anisotropic scattering treatment is required.
Date: December 31, 1991
Creator: Hill, R. N.; Fanning, T. H. & Finck, P. J.
Partner: UNT Libraries Government Documents Department

IAEA sodium void reactivity benchmark calculations

Description: In this paper, the IAEA-1 992 ``Benchmark Calculation of Sodium Void Reactivity Effect in Fast Reactor Core`` problem is evaluated. The proposed design is a large axially heterogeneous oxide-fueled fast reactor as described in Section 2; the core utilizes a sodium plenum above the core to enhance leakage effects. The calculation methods used in this benchmark evaluation are described in Section 3. In Section 4, the calculated core performance results for the benchmark reactor model are presented; and in Section 5, the influence of steel and interstitial sodium heterogeneity effects is estimated.
Date: December 1, 1992
Creator: Hill, R. N. & Finck, P. J.
Partner: UNT Libraries Government Documents Department