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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

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

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

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

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

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

Annular Core Pulse Reactor upgrade quarterly report, January--March 1977

Description: Information is presented concerning safety, compliance, and documentation; core nuclear design; ACPR upgrade and console development; mechanical design; fuel element design; fuel element fabrication; secondary fuel materials studies; driver core fuel element; and diagnostic system.
Date: June 1, 1977
Creator: Walker, J. V.
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

Calculated physics and performance parameters for the ACPR upgrade

Description: Reactor physics and core design calculations have been completed for the upgrade of the Annular Core Pulse Reactor (ACPR) at Sandia Laboratories. The ACPR has been in operation since 1967 and is a U-Zr H/sub 1.6/ fueled swimming pool type reactor with a large dry central experiment cavity. The purpose of the upgrade of the ACPR is to provide improved pulse and steady-state capabilities for performing neutron simulation and reactor safety tests. The performance goals for the ACPR Upgrade were to achieve a factor of 2.6 times the pulse fluence available with the ACPR and a factor of 2 increase in the steady-state flux. In addition, the pulse width of the upgraded reactor was to be as short as practical to maintain the rapid transient simulation capability of the ACPR.
Date: January 1, 1977
Creator: Pickard, P. S. & Odom, J. P.
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

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

Interim reliability evaluation program, Browns Ferry fault trees

Description: An abbreviated fault tree method is used to evaluate and model Browns Ferry systems in the Interim Reliability Evaluation programs, simplifying the recording and displaying of events, yet maintaining the system of identifying faults. The level of investigation is not changed. The analytical thought process inherent in the conventional method is not compromised. But the abbreviated method takes less time, and the fault modes are much more visible.
Date: January 1, 1981
Creator: Stewart, M.E.
Partner: UNT Libraries Government Documents Department

Explosive demolition of activated concrete

Description: This paper describes the removal of a radiologically contaminated concrete pad. This pad was removed during 1979 by operating personnel under the direction of the Waste Management Program of EG and G Idaho, Inc. The concrete pad was the foundation for the Organic Moderated Reactor Experiment (OMRE) reactor vessel located at the Idaho National Engineering Laboratory (INEL). The pad consisted of a cylindrical concrete slab 15 ft in diameter, 2 ft thick, and reinforced with steel bar. It was poured directly onto basalt rocks approximately 20 ft below grade. The entire pad contained induced radioactivity and was therefore demolished, boxed, and buried rather than being decontaminated. The pad was demolished by explosive blasting.
Date: January 1, 1980
Creator: Smith, D.L.
Partner: UNT Libraries Government Documents Department