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Physics and feasibility study of the Fast-Mixed Spectrum Reactor concept

Description: Reactor physics and fuel cycle studies, coordinated with heat transfer and material science and structural analysis work has indicated the feasibility potential of the coupled Fast-Mixed Spectrum Reactor (FMSR) concept. This concept employs what are considered reasonable extrapolations of present fast breeder reactor technology to achieve a once-through-and-store reactor fuel cycle. Since the fuel cycle for this reactor is intended to use only natural or depleted uranium for its equilibrium feed, the resultant reactor would have excellent anti-proliferation characteristics. It would also extend utilization of natural uranium resources by a factor of about 15 relative to LWR reactors when on its equilibrium fuel cycle; startup requirements would of course reduce this factor.
Date: January 1, 1979
Creator: Fischer, G.J.; Kouts, H.J.C.; Cerbone, R.J.; Shenoy, S.; Durston, C.; Ludewig, H. et al.
Partner: UNT Libraries Government Documents Department

Fast mixed spectrum reactor concept

Description: The Fast Mixed Spectrum Reactor is a highly promising concept for a fast reactor with improved features of proliferation resistance, and excellent utilization of uranium resources. In technology, it can be considered to be a branch of fast breeder development, though its operation and implications are different from those of FBR'S in important respects. Successful development programs are required in several areas to bring FMSR to reality, but the payoff from a successful program can be high.
Date: April 1, 1979
Creator: Kouts, H.J.C.; Fischer, G.J. & Cerbone, R.J.
Partner: UNT Libraries Government Documents Department

Analysis of fission-product effects in a Fast Mixed-Spectrum Reactor concept

Description: The Fast Mixed-Spectrum Reactor (FMSR) concept has been proposed by BNL as a means of alleviating certain nonproliferation concerns relating to civilian nuclear power. This breeder reactor concept has been tailored to operate on natural uranium feed (after initial startup), thus eliminating the need for fuel reprocessing. The fissile material required for criticality is produced, in situ, from the fertile feed material. This process requires that large burnup and fluence levels be achievable, which, in turn, necessarily implies that large fission-product inventories will exist in the reactor. It was the purpose of this study to investigate the effects of large fission-product inventories and to analyze the effect of burnup on fission-product nuclide distributions and effective cross sections. In addition, BNL requested that a representative 50-group fission-product library be generated for use in FMSR design calculations.
Date: February 1, 1980
Creator: White, J.R. & Burns, T.J.
Partner: UNT Libraries Government Documents Department

Two-lump fission product model for fast reactor analysis

Description: As a part of the Fast-Mixed Spectrum Reactor (FMSR) Project, a study was made on the adequacy of the conventional fission product lump models for the analysis of the different FMSR core concepts. A two-lump fission product model consisting of an odd-A fission product lump and an even-A fission product lump with transmutation between the odd- and even-A lumps was developed. This two-lump model is capable of predicting the exact burnup-dependent behavior of the fission products within a few percent over a wide range of spectra and is therefore also applicable to the conventional fast breeder reactor.
Date: January 1, 1981
Creator: Atefi, B
Partner: UNT Libraries Government Documents Department

Fast-Mixed Spectrum Reactor progress report. Results of the FMSR Benchmark calculations and an assessment of current fission product libraries

Description: As part of the Initial Feasibility Study of the Fast Mixed Spectrum Reactor, a series of benchmark calculations were made to determine the sensitivity of the physics analysis to differences in methods and data. Argonne National Laboratory (ANL), the Massachusetts Institute of Technology (MIT), and Oak Ridge National Laboratory (ORNL) were invited to participate with Brookhaven National Laboratory in the analysis of a FMSR model prescribed by BNL. Detailed comparisons are made including a comprehensive study on the adequacy of the fission product treatments.
Date: June 1, 1980
Creator: Ludewig, H.; Durston, C.; Atefi, B. & Cerbone, R.J.
Partner: UNT Libraries Government Documents Department

Human factors review for nuclear power plant severe accident sequence analysis

Description: The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function-oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage preventing radiological release.
Date: January 1, 1985
Creator: Krois, P.A. & Haas, P.M.
Partner: UNT Libraries Government Documents Department

Importance of momentum dynamics in BWR neutronic stability: experimental evidence

Description: Momentum dynamics affect the boiling water reactor (BWR) neutronic stability by coupling steam void perturbations and core-inlet coolant flow. Computer simulations have shown that proper modeling of the recirculation loop, which shares the upper and lower plena pressures with the reactor core, is essential for accurate stability calculations. Purpose of this paper is to show experimental evidence, obtained from a recent series of stability tests performed at the Browns Ferry-1 BWR, demonstrating the important role of momentum dynamics in BWR neutronic stability.
Date: January 1, 1985
Creator: March-Leuba, J. & Otaduy, P.J.
Partner: UNT Libraries Government Documents Department

Interim reliability evaluation program, Browns Ferry 1

Description: Probabilistic risk analysis techniques, i.e., event tree and fault tree analysis, were utilized to provide a risk assessment of the Browns Ferry Nuclear Plant Unit 1. Browns Ferry 1 is a General Electric boiling water reactor of the BWR 4 product line with a Mark 1 (drywell and torus) containment. Within the guidelines of the IREP Procedure and Schedule Guide, dominant accident sequences that contribute to public health and safety risks were identified and grouped according to release categories.
Date: January 1, 1981
Creator: Mays, S.E.; Poloski, J.P.; Sullivan, W.H.; Trainer, J.E.; Bertucio, R.C. & Leahy, T.J.
Partner: UNT Libraries Government Documents Department

Design and proposed utilization of the Sandia Annular Core Research Reactor (ACRR)

Description: The Sandia ACRR became operational in 1978 and currently serves as the major in-pile fast reactor safety test facility for the US Nuclear Regulatory Commission. The ACRR is an upgrade of the Annular Core Pulse Reactor (ACPR) with the installation of a new flexible control system and a core of uniquely designed BeO-UO/sub 2/ fuel elements for increasing the neutron fluence in the experiment cavity. The reactor is now capable of driving multi-pin advanced reactor test fuel into vapor with a pulse width of 5 msec. In the steady state mode, the reactor can simulate post accident decay heat at prototypic levels in fission heated debris beds up to 10 cm in diameter. A number of programmed operating modes including high power square waves, ramps and pulses can produce a multitude of power profiles in order to simulate the power histories in the various accident scenarios. The reactor capabilities and the reactor safety research test program are discussed.
Date: January 1, 1979
Creator: Walker, J.V.; Reuscher, J.A. & Pickard, P.S.
Partner: UNT Libraries Government Documents Department

Severe-accident-sequence assessment of hypothetical complete-station blackout at the Browns Ferry Nuclear Plant

Description: An investigation has been made of various accident sequence which may occur following a complete loss of offsite and onsite ac power at a Boiling Water Reactor nuclear power plant. The investigation was performed for the Browns Ferry Nuclear Power Plant, and all accident sequences resulted in a hypothetical core meltdown. Detailed calculations were performed with the MARCH computer meltdown. Detailed calcuations were performed with the MARCH computer code containing a decay power calculation which was modified to include the actinides. This change has resulted in shortening the time before core uncovery by approx. 18%, and reducing the time before the start of core melting by approx. 26%. Following the hypothetical core meltdown accident, the drywell electric penetration assembly seals have been identified as the most likely leak pathway outside the containment. This potential mode of containment failure occurs at a pressure approx. 30% lower than that analyzed in the Reactor Safety Study.
Date: January 1, 1981
Creator: Yue, D.D. & Condon, W.A.
Partner: UNT Libraries Government Documents Department

RAMONA-3B application to Browns Ferry ATWS

Description: This paper discusses two preliminary MSIV clsoure ATWS calculations done using the RAMONA-3B code and the work being done to create the necessary cross section sets for the Browns Ferry Unit 1 reactor. The RAMONA-3B code employs a three-dimensional neutron kinetics model coupled with one-dimensional, four equation, nonhomogeneous, nonequilibrium thermal hydraulics. To be compatible with 3-D neutron kinetics, the code uses parallel coolant channels in the core. It also includes a boron transport model and all necessary BWR components such as jet pump, recirculation pump, steam separator, steamline with safety and relief valves, main steam isolation valve, turbine stop valve, and turbine bypass valve. A summary of RAMONA-3B neutron kinetics and thermal hydraulics models is presented in the Appendix.
Date: January 1, 1984
Creator: Slovik, G.C.; Neymotin, L.; Cazzoli, E. & Saha, P.
Partner: UNT Libraries Government Documents Department

Fast-mixed spectrum reactor. Progress report for 1980

Description: Reactor physics, fuel cycle, thermal-hydraulics and fuel cycle cost studies have been performed for this concept and are reported. The most serious drawback of previous FMSR designs, namely the level of irradiation damage to the stainless steel of the cladding and duct materials, has been greatly reduced by the new design. The peak fuel burnup level is also reduced. Work continued on earlier FMSR designs, and in particular, the centrally-moderated FMSR. Emphasis was placed on defining the first core and then the total sequence of core histories over the 30-year life of the reactor. It was found possible to define a two-year fuel cycle with limited reactivity swing over the cycle. Fuel cycle cost studies were begun. The results indicate a modest fuel cycle cost advantage for the FMSR, but the basic cost assumptions must be improved for metal fuel. Improved thermal-hydraulic analysis capabilities have greatly improved the understanding of heat transfer behavior.
Date: October 1, 1980
Creator: Fischer, G.J.; Galperin, A.; Shenoy, S. & Atefi, B.
Partner: UNT Libraries Government Documents Department

Potential effects of the fire protection system sprays at Browns Ferry on fission product transport

Description: The fire protection system (FPS) sprays within any nuclear plant are not intended to mitigate radioactive releases to the environment resulting from severe core-damage accidents. However, it has been shown here that during certain postulated severe accident scenarios at the Browns Ferry Nuclear Plant, the functioning of FPS sprays could have a significant impact on the radioactive releases. Thus the effects of those sprays need to be taken into account for realistic estimation of source terms for some accident scenarios. The effects would include direct ones such as cooling of the reactor building atmosphere and scrubbing of radioactivity from it, as well as indirect effects such as an altered likelihood of hydrogen burning and flooding of various safety-related pumps in the reactor building basement. Thus some of the impacts of the sprays would be beneficial with respect to mitigating releases to the environment but some others might not be. The effects of the FPS would be very scenario dependent with a wide range of potential effects often existing for a given accident sequence. Any generalization of the specific results presented here for Browns Ferry to other nuclear plants must be done cautiously, as it appears from a preliminary investigation that the relevant physical and operational characteristics of FPS spray systems differ widely among even otherwise apparently similar plants. Likewise the standby gas treatment systems, which substantially impact the effects of the FPS, differ significantly among plants. More work for both Mark I plants and other plants, BWRs and PWRs alike, is indicated so the potential effects of FPS spray systems during severe accidents can be at least ball-parked for more realistic accident analyses.
Date: January 1, 1983
Creator: Niemczyk, S.J.
Partner: UNT Libraries Government Documents Department

New neutron simulation capabilities provided by the Sandia Pulse Reactor (SPR-III) and the Upgraded Annular Core Pulse Reactor (ACPR)

Description: The paper briefly describes the nuclear reactor facilities at Sandia Laboratories which are used for simulating nuclear weapon produced neutron environments. These reactor facilities are used principally in support of continuing R and D programs for the Department of Energy/Office of Military Application (DOE/OMA) in studying the effects of radiation on nuclear weapon systems and components. As such, the reactors are available to DOE and DOD agencies and their contractors responsible for the radiation hardening of advanced nuclear weapon systems. Emphasis is placed upon two new reactor simulation sources; the Sandia Pulse Reactor-III (SPR-III) Facility which enhances the neutron exposure volume capabilities over those presently available with the existing SPR-II Facility, and the Upgraded Annular Core Pulse Reactor (ACPR) Facility which enhances the neutron exposure capabilities over those of the former ACPR Facility.
Date: July 1, 1978
Creator: Choate, L.M. & Schmidt, T.R.
Partner: UNT Libraries Government Documents Department

Remote encapsulation of mixed-oxide fuel pellets for transient testing

Description: Nine individual fuel pellets from irradiated fuel pins were extracted and encapsulated remotely in double containment canisters in support of the simulated loss-of-coolant flow tests being conducted in the Sandia Laboratory Annular Core Pulsed Reactor.
Date: January 1, 1978
Creator: Dowler, K.E.; Newbury, F.H.; Ledbetter, J.M. & Cano, G.L.
Partner: UNT Libraries Government Documents Department

Fast-Mixed Spectrum Reactor. Progress report for 1979

Description: This report summarizes the progress of the Fast Mixed Spectrum Reactor (FMSR) since the publication of the Interim Report in January 1979. The FMSR program was initiated to determine the feasibility of a breeder reactor concept which operated on a once-through-and-store fuel cycle and for which the only feed would be natural uranium. A first or startup core enriched to a maximum of about eleven percent in uranium-235 would be required. The concept has excellent antiproliferation advantages. In the once-through and store mode, the FMSR has a resource utilization which is a factor of four higher than a light water reactor.
Date: May 1, 1980
Creator: Fischer, G.J. & Cerbone, R.J.
Partner: UNT Libraries Government Documents Department

Fuel cycle analysis of once-through nuclear systems.

Description: Once-through fuel cycle systems are commercially used for the generation of nuclear power, with little exception. The bulk of these once-through systems have been water-cooled reactors (light-water and heavy water reactors, LWRs and HWRs). Some gas-cooled reactors are used in the United Kingdom. The commercial power systems that are exceptions use limited recycle (currently one recycle) of transuranic elements, primarily plutonium, as done in Europe and nearing deployment in Japan. For most of these once-through fuel cycles, the ultimate storage of the used (spent) nuclear fuel (UNF, SNF) will be in a geologic repository. Besides the commercial nuclear plants, new once-through concepts are being proposed for various objectives under international advanced nuclear fuel cycle studies and by industrial and venture capital groups. Some of the objectives for these systems include: (1) Long life core for remote use or foreign export and to support proliferation risk reduction goals - In these systems the intent is to achieve very long core-life with no refueling and limited or no access to the fuel. Most of these systems are fast spectrum systems and have been designed with the intent to improve plant economics, minimize nuclear waste, enhance system safety, and reduce proliferation risk. Some of these designs are being developed under Generation IV International Forum activities and have generally not used fuel blankets and have limited the fissile content of the fuel to less than 20% for the purpose on meeting international nonproliferation objectives. In general, the systems attempt to use transuranic elements (TRU) produced in current commercial nuclear power plants as this is seen as a way to minimize the amount of the problematic radio-nuclides that have to be stored in a repository. In this case, however, the reprocessing of the commercial LWR UNF to produce the initial fuel will be necessary. For ...
Date: August 10, 2010
Creator: Kim, T. K.; Taiwo, T. A. & Division, Nuclear Engineering
Partner: UNT Libraries Government Documents Department

Theory and application of the coded aperture fuel motion detection system

Description: A fuel motion detection system based on coded aperture imaging has been developed for the Annular Core Research Reactor. Its configuration evolved after investigations were carried out to determine the required system capabilities. The reactor environment, developments in the theory of coded apertures for nuclear radiations and compatibility with prototypical geometries were considered. The system was fabricated and inserted into the ACRR where it has recorded the fuel motion from a single pin subjected to loss of flow accident conditions. In addition, computer simulations have shown that in the reactor environment and for fast data acquisition, coded imaging, particularly with uniformly redundant arrays, offers significant advantages over pinhole camera geometries. The extension of this technique toward imaging of 37-pin bundles and imaging with fast neutrons is also being investigated.
Date: January 1, 1979
Creator: Kelly, J.G.; Stalker, K.T.; McArthur, D.A.; Chu, K.W. & Powell, J.E.
Partner: UNT Libraries Government Documents Department

BWR severe accident sequence analyses at ORNL - some lessons learned

Description: Boiling water reactor severe accident sequence studies are being carried out using Browns Ferry Unit 1 as the model plant. Four accident studies were completed, resulting in recommendations for improvements in system design, emergency procedures, and operator training. Computer code improvements were an important by-product.
Date: January 1, 1983
Creator: Hodge, S.A.
Partner: UNT Libraries Government Documents Department

Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)

Description: Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs., 4 figs., 1 tab.
Date: January 1, 1991
Creator: Ball, R.M.; Madaras, J.J. (B and W Nuclear Technologies, Lynchburg, VA (USA). Space and Defense Systems); Trowbridge, F.R. Jr.; Talley, D.G. & Parma, E.J. Jr. (Sandia National Labs., Albuquerque, NM (USA))
Partner: UNT Libraries Government Documents Department

D5 debris bed experiment-extended post-dryout observations of a UO/sub 2/-sodium particle bed. [In ACPR reactor]

Description: The D5 experiment is the sixth in a series of debris bed coolability experiments. The principal objective of the experiment is to investigate debris bed behavior, for various sodium subcoolings, at temperatures exceeding 2100/sup 0/K. The experiment is scheduled for early February 1982.
Date: January 1, 1982
Creator: Gronager, J E
Partner: UNT Libraries Government Documents Department

D9 experiment: heat removal from stratified UO/sub 2/ debris

Description: The D9 experiment investigated the coolability of a shallow (77 mm), stratified urania bed in sodium. The bed was fission heated in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories to simulate the effects of radioactive decay heating. It was the first stratified debris bed experiment to use an extended UO/sub 2/ particle size distribution (0.038 to 4.0 mm). Dryout occurred at powers ranging from 0.10 to 0.58 W/g, which was close to the incipient boiling power and before channels penetrated the subcooled zone in the bed, even with subcoolings as low as 80/sup 0/C. Channel penetration was observed after dryout began, but the bed became only moderately more coolable. All these observations agree with current models.
Date: April 1, 1985
Creator: Ottinger, C A; Mitchell, G W; Lipinski, R J & Kelly, J E
Partner: UNT Libraries Government Documents Department

D7 debris bed experiment - heat removal from a shallow, stratified UO/sub 2/-sodium particle bed. [In ACPR reactor]

Description: The D7 debris bed experiment is one in a series of experiments being conducted to investigate the coolability of fragmented reactor mterials. The D7 experiment is the second experiment which investigates heat removal from a stratified bed. The significance of particle stratification was demonstrated by the D6 experiment, which exhibited lower dryout powers of a factor of approximately two when compared to mixed beds. In addition to investigating a shallower bed as compared to the D6 experiment, the D7 experiment investigates a wider range of sodium subcooling achievable during the experiment.
Date: January 1, 1982
Creator: Mitchell, G W
Partner: UNT Libraries Government Documents Department

Analysis of loss of decay-heat-removal sequences at Browns Ferry Unit One

Description: This paper summarizes the Oak Ridge National Laboratory (ORNL) report Loss of DHR Sequences at Browns Ferry Unit One - Accident Sequence Analysis (NUREG/CR-2973). The Loss of DHR investigation is the third in a series of accident studies concerning the BWR 4 - MK I containment plant design. These studies, sponsored by the Nuclear Regulatory Commission Severe Accident Sequence Analysis (SASA) program, have been conducted at ORNL with the full cooperation of the Tennessee Valley Authority (TVA). The purpose of the SASA studies is to predetermine the probable course of postulated severe accidents so as to establish the timing and the sequence of events. The SASA studies also produce recommendations concerning the implementation of better system design and better emergency operating instructions and operator training. The ORNL studies also include a detailed, best-estimate calculation of the release and transport of radioactive fission products following postulated severe accidents.
Date: January 1, 1983
Creator: Harrington, R.M.
Partner: UNT Libraries Government Documents Department