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A technique for computing bowing reactivity feedback in LMFBR's

Description: During normal or accidental transients occurring in a LMFBR core, the assemblies and their support structure are subjected to important thermal gradients which induce differential thermal expansions of the walls of the hexcans and differential displacement of the assembly support structure. These displacements, combined with the creep and swelling of structural materials, remain quite small, but the resulting reactivity changes constitute a significant component of the reactivity feedback coefficients used in safety analyses. It would be prohibitive to compute the reactivity changes due to all transients. Thus, the usual practice is to generate reactivity gradient tables. The purpose of the work presented here is twofold: develop and validate an efficient and accurate scheme for computing these reactivity tables; and to qualify this scheme.
Date: January 1, 1987
Creator: Finck, P.J.
Partner: UNT Libraries Government Documents Department

Compared performances of ENDF/B-VI and JEF-2.2 for MOX core physics.

Description: The US is currently evaluating the use of MOX fuel in commercial LWR's for reducing weapons grade Pu stockpiles. The design and licensing processes will require that the validity of the nuclear data libraries and codes used in the effort be demonstrated. Unfortunately, there are only a very limited number of relatively old and non representative integral experiments' freely available to the US programs. This lack of adequate experimental data can be partially remediated by comparing the results of well validated European codes with the results of candidate US codes. The demonstration can actually be divided in two components: a code to code (Monte Carlo) comparison can easily demonstrate the validity and limits of the proposed algorithms; and the performances of nuclear data libraries should be compared, major trends should be observed, and their origins should be explained in terms of differences in evaluated nuclear data; In this paper, we have compared the performances of the JEF-2.2 and ENDF/B-VI.4 libraries for a series of benchmarks for k{sub eff}, void worth, and pin power distributions. Note that JEF-2.2 has been extensively validated for MOX applications.
Date: July 8, 1998
Creator: Finck, P. J.
Partner: UNT Libraries Government Documents Department

The Physics of transmutation systems : system capabilities and performances.

Description: This document is complementary to a document produced by Prof. Salvatores on ''The Physics of Transmutation in Critical or Subcritical Reactors and the Impact on the Fuel Cycle''. In that document, Salvatores describes the fundamental of transmutation, through basic physics properties and general parametric studies. In the present document we try to go one step further towards practical implementation (while recognizing that the practical issues such as technology development and demonstration, and economics, can only be mentioned in a very superficial manner). Section 1 briefly overviews the possible objectives of transmutation systems, and links these different objectives to possible technological paths. It also describes the overall constraints which have to be considered when developing and implementing transmutation systems. In section 2 we briefly overview the technological constraints which need to be accounted for when designing transmutation systems. In section 3 we attempt to provide a simplified classification of transmutation systems in order to clarify later comparisons. It compares heterogeneous and homogeneous recycle strategies, and single and multi-tier systems. Section 4 presents case analyses for assessing the transmutation performance of various individual systems, starting with LWR's (1. generic results; 2. multirecycle of plutonium; 3. an alternative: transmutation based on a Thorium fuel cycle), followed by Gas-Cooled Reactors (with an emphasis on the ''deep burn'' approach), and followed by Fast Reactors and Accelerator Driven systems (1. generic results; 2. homogeneous recycle of transuranics; 3. practical limit between Fast Reactors and Accelerator Driven Systems) Section 5 summarizes recent results on integrated system performances. It focuses first on interface effects between the two elements of a dual tier system, and then summarizes the major lessons learned from recent global physics studies.
Date: August 21, 2002
Creator: 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: November 1992
Creator: Hill, R. N. & 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

Assessment of General Atomics accelerator transmutation of waste concept based on gas-turbine-modular helium cooled reactor technology.

Description: An assessment has been performed for an Accelerator Transmutation of Waste (ATW) concept based on the use of the high temperature gas reactor technology. The concept has been proposed by General Atomics for the ATW system. The assessment was jointly conducted at Argonne National Laboratory (ANL) and Los Alamos national laboratory to assess and to define the potential candidates for the ATW system. This report represents the assessment work performed at ANL. The concept uses recycled light water reactor (LWR)-discharge-transuranic extracted from irradiated oxide fuel in a critical and sub-critical accelerator driven gas-cooled transmuter. In this concept, the transmuter operates at 600 MWt first in the critical mode for three cycles and then operates in a subcritical accelerator-driven mode for a single cycle. The transmuter contains both thermal and fast spectrum transmutation zones. The thermal zone is fueled with the TRU oxide material in the form of coated particles, which are mixed with graphite powder, packed into cylindrical compacts, and loaded in hexagonal graphite blocks with cylindrical channels; the fast zone is fueled with TRU-oxide material in the form of coated particles without the graphite powder and the graphite blocks that has been burned in the thermal region for three critical cycles and one additional accelerator-driven cycle. The fuel loaded into the fast zone is irradiated for four additional cycles. This fuel management scheme is intended to achieve a high Pu isotopes consumption in the thermal spectrum zone, and to consume the minor actinides in the fast-spectrum zone. Monte Carlo and deterministic codes have been used to assess the system performance and to determine the feasibility of achieving high TRU consumption levels. The studies revealed the potential for high consumption of Pu-239 (97%), total Pu (71%) and total TRU (64%) in the system. The analyses confirmed the need for burnable ...
Date: May 8, 2001
Creator: Gohar, Y.; Taiwo, T. A.; Cahalan, J. E. & Finck, P. J.
Partner: UNT Libraries Government Documents Department

Assessment of the teledial gas-cooled transmuter concept

Description: The application of four gas-turbine, modular helium cooled reactors and an accelerator unit (GT/AD-MHR) has been proposed for burning transuranics recycled from LWR waste. The recycled LWR discharged transuranics encapsulated in TRISO coated particles are first loaded into the outer thermal spectrum zone of the GT/AD-MHR for burning in the critical mode for about three years. Previously burned fuel is in a central fast zone. In the fourth year, the same unit is configured as an accelerator-driven system, containing a centrally located spallation target. The three-year, thermal-zone burned fuel and the inner fast-zone fuel from the critical mode operation are used in this subcritical cycle, and remain in their respective zones. At the end of this one-year subcritical irradiation, the outer thermal-zone fuel is reconstituted and used as fast-zone fuel in another critical mode operation. As the fuel in the fast-zone has reached its end of life it is discharged, with very low transuranics content. The critical mode operation is staggered, and each GT/AD-MGR unit undergoes the subcritical burn in one out of four year. The physics performance of the GT/AD-MHR has been evaluated using independent deterministic and Monte Carlo codes and the results of the study are presented in the current paper. A companion paper discussing the verification of the codes is also presented at this meeting. Single-batch and three-batch fuel loading schemes for the GT/AD-MHR have been evaluated using the REBUS-3/DIF3D fuel cycle code, to determine the feasibility of achieving very high burnup without exceeding reactivity and power density limits. The reactor physics of the GT-MHR is complicated by the presence of the low-lying plutonium and Er-167 resonances (0.2--1.1 eV) and by the fact that the neutron spectrum has a low-energy peak about this energy range. This peak can change depending on the core state or material loading. ...
Date: July 24, 2000
Creator: Taiwo, T. A.; Gohar, Y. & Finck, P. J.
Partner: UNT Libraries Government Documents Department

Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.

Description: Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.
Date: February 17, 1999
Creator: Blomquist, R.N.; Finck, P.J.; Jammes, C. & Stenberg, C.G.
Partner: UNT Libraries Government Documents Department

Uncertainty assessment for accelerator-driven systems.

Description: The concept of a subcritical system driven by an external source of neutrons provided by an accelerator ADS (Accelerator Driver System) has been recently revived and is becoming more popular in the world technical community with active programs in Europe, Russia, Japan, and the U.S. A general consensus has been reached in adopting for the subcritical component a fast spectrum liquid metal cooled configuration. Both a lead-bismuth eutectic, sodium and gas are being considered as a coolant; each has advantages and disadvantages. The major expected advantage is that subcriticality avoids reactivity induced transients. The potentially large subcriticality margin also should allow for the introduction of very significant quantities of waste products (minor Actinides and Fission Products) which negatively impact the safety characteristics of standard cores. In the U.S. these arguments are the basis for the development of the Accelerator Transmutation of Waste (ATW), which has significant potential in reducing nuclear waste levels. Up to now, neutronic calculations have not attached uncertainties on the values of the main nuclear integral parameters that characterize the system. Many of these parameters (e.g., degree of subcriticality) are crucial to demonstrate the validity and feasibility of this concept. In this paper we will consider uncertainties related to nuclear data only. The present knowledge of the cross sections of many isotopes that are not usually utilized in existing reactors (like Bi, Pb-207, Pb-208, and also Minor Actinides and Fission Products) suggests that uncertainties in the integral parameters will be significantly larger than for conventional reactor systems, and this raises concerns on the neutronic performance of those systems.
Date: June 10, 1999
Creator: Finck, P. J.; Gomes, I.; Micklich, B. & Palmiotti, G.
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

Analysis of EBR-II neutron and photon physics by multidimensional transport-theory techniques

Description: This paper contains a review of the challenges specific to the EBR-II core physics, a description of the methods and techniques which have been developed for addressing these challenges, and the results of some validation studies relative to power-distribution calculations. Numerical tests have shown that the VARIANT nodal code yields eigenvalue and power predictions as accurate as finite difference and discrete ordinates transport codes, at a small fraction of the cost. Comparisons with continuous-energy Monte Carlo results have proven that the errors introduced by the use of the diffusion-theory approximation in the collapsing procedure to obtain broad-group cross sections, kerma factors, and photon-production matrices, have a small impact on the EBR-II neutron/photon power distribution.
Date: March 1, 1994
Creator: Jacqmin, R. P.; Finck, P. J. & Palmiotti, G.
Partner: UNT Libraries Government Documents Department

Lead-bismuth target design for the subcritical multiplier (SCM) of the accelerator driven test facility (ADTF).

Description: A lead-bismuth eutectic (LBE) target design concept has been developed to drive the subcritical multiplier (SCM) of the accelerator-driven test facility (ADTF). This report gives the target design description, the results from the parametric studies, and the design analyses including physics, heat-transfer, hydraulics, structural, radiological, and safety analyses. The design is based on a coaxial geometrical configuration to minimize the target footprint and to maximize the utilization of the spallation neutrons. The target is installed vertically along the SCM axis. LBE is the target material and the target coolant. Ferritic steel (HT-9 alloy) is the selected structural material based on the current database and the design analyses. Austenitic steel (Type 316 stainless steel) is the backup choice. A uniform proton beam is employed to perform the spallation process. The proton beam has 8.33-mA current and 8.14-cm radius resulting in a current density of 40 {micro}A/cm{sup 2}. The beam power is 5 MW and the proton energy is 600 MeV. The beam tube has 10-cm radius to accommodate the halo current. A hemi-spherical geometry is used for the target window, which is connected to the beam tube. A conical target window with a rounded tip is also considered since it has a lower average temperature relative to the spherical geometry. The beam tube is enclosed inside two coaxial tubes to provide inlet and outlet manifolds for the LBE coolant. The inlet and the outlet coolant manifolds and the proton beam are entered from the top above the SCM. Several design constraints are developed and utilized for the target design process to satisfy different engineering requirements and to minimize the design development time and cost.
Date: April 18, 2002
Creator: Gohar, Y.; Finck, P. J.; Krajtl, L.; Herceg, J.; Pointer, W. D.; Saiveau, J. et al.
Partner: UNT Libraries Government Documents Department

Advanced burner test reactor preconceptual design report.

Description: The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing ...
Date: December 16, 2008
Creator: Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F. et al.
Partner: UNT Libraries Government Documents Department