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Neutron Flux and Spectra Measurements in the Sandia Pulsed Reactor Facility (SPRF)

Description: Introduction: Neutron measurements were made in the pulsed reactor building and on the safety screen of the pulsed reactor in order to determine the neutron yield of the reactor as a function of (1) distance from the reactor centerline, (2) direction in the reactor building, and (3) position on the reactor safety screen.
Date: January 1962
Creator: Buckalew, William H.
Partner: UNT Libraries Government Documents Department

Pulsed reactor experiments at Oak Ridge

Description: This paper describes dynamic experiments for 3 pulsed reactors. 1st reactor was pulsed from some average power by rotating a partial Be reflector past an unreflected core face; the other 2 reactors were pulsed by rapid insertion of a fuel rod into the unmoderated and unreflected reactor at essentially zero neutron level with no significant neutron source present. The first reactor was a mockup of an EURATOM design (never constructed) of the proposed SORA Reactor, and the other two were the Health Physics Research Reactor and the Army Pulse Radiation Facility Reactor (APRFR). This paper describes the experiments performed in initial testing of these systems, including destructive tests of APRFR, to set operating limits for this type of reactor in pulsed operation. All the experiments described were performed at the Oak Ridge Critical Experiments Facility.
Date: December 31, 1994
Creator: Mihalczo, J.T.
Partner: UNT Libraries Government Documents Department

Homopolar Gun for Pulsed Spheromak Fusion Reactors II

Description: A homopolar gun is discussed that could produce the high currents required for pulsed spheromak fusion reactors even with unit current amplification and open field lines during injection, possible because close coupling between the gun and flux conserver reduces gun losses to acceptable levels. Example parameters are given for a gun compatible with low cost pulsed reactors and for experiments to develop the concept.
Date: June 14, 2004
Creator: Fowler, T
Partner: UNT Libraries Government Documents Department

Overview of Sandia National Laboratories pulse nuclear reactors

Description: Sandia National Laboratories has designed, constructed and operated bare metal Godiva-type and pool-type pulse reactors since 1961. The reactor facilities were designed to support a wide spectrum of research, development, and testing activities associated with weapon and reactor systems.
Date: October 1, 1994
Creator: Schmidt, T. R. & Reuscher, J. A.
Partner: UNT Libraries Government Documents Department

The shift of prompt critical in reflected reactors and the limitations of the mean prompt-neutron lifetime model

Description: Prompt critical in a bare reactor is defined as the point at which the reactivity {rho} of the reactor is equal to the effective delayed neutron fraction {beta}. In a reflected reactor, however, it is shown that prompt critical will occur at a reactivity of {rho} = {beta}(1-f) where f is the fraction of core neutrons that return to the core region after having leaked into the reflector. Furthermore, it is also shown that the mean prompt-neutron lifetime model that has been traditionally used to characterize the dynamic response of reflected reactors may not always provide an adequate representation of the system for reactivities greater than 1$. And finally, the coupled, point-kinetic equations proposed by Avery and further developed by Cohn for simple reflected systems are recast into a more usable form that can be readily used to perform superprompt critical transient analyses.
Date: August 1, 1994
Creator: Spriggs, G. D. & Busch, R. D.
Partner: UNT Libraries Government Documents Department

U.S./Russian cooperative efforts to enhance nuclear MPC&A at VNIITF, (Chelyabinsk-70)

Description: The work described here is part of an effort called the Nuclear Materials Protection, Control, and Accounting (MPC&A) Program, a cooperative program between the US Department of Eenrgy (DOE) and Russia's Ministry of Atomic Energy (MinAtom). The objective of the program is to reduce the risk of nuclear proliferation by strengthening MPC&A systems at Russian nuclear Facilities. This paper describes that portion of the MPC&A program that is directed specifically to the needs of the All Russian Scientific Research Institute of Technical Physics (VNIITF), also called Chelyabinsk-70. A major MPC&A milestone was met at VNIITF when the MPC&A improvements were commissioned at the Pulse Research Reactor Facility in May of this year.
Date: April 20, 1999
Creator: Abramson, B; Apt, K; Blasy, J; Bukin, D; Churikov, Y; Curtis, D et al.
Partner: UNT Libraries Government Documents Department

Godiva IV and Juliet Diagnostics CED-1, Rev. 1 (IER-176)

Description: The Juliet experiment is currently in preliminary design (IER-128). This experiment will utilize a suite of diagnostics to measure the physical state of the device (temperature, surface motion, stress, etc.) and the total and time rate of change of neutron and gamma fluxes. A variety of potential diagnostics has been proposed in this CED-1 report. Based on schedule and funding, a subset of diagnostics will be selected for testing using the Godiva IV pulsed reactor as a source of neutrons and gammas. The diagnostics development and testing will occur over a two year period (FY12-13) culminating in a final set of diagnostics to be fielded for he Juliet experiment currently proposed for execution in FY15.
Date: April 11, 2012
Creator: Scorby, J C & Myers, W L
Partner: UNT Libraries Government Documents Department

Photoneutron effects on pulse reactor kinetics for the Annular Core Research Reactor (ACRR).

Description: The Annular Core Research Reactor (ACRR) is a swimming-pool type pulsed reactor that maintains an epithermal neutron flux and a nine-inch diameter central dry cavity. One of its uses is neutron and gamma-ray irradiation damage studies on electronic components under transient reactor power conditions. In analyzing the experimental results, careful attention must be paid to the kinetics associated with the reactor to ensure that the transient behavior of the electronic device is understood. Since the ACRR fuel maintains a substantial amount of beryllium, copious quantities of photoneutrons are produced that can significantly alter the expected behavior of the reactor power, especially following a reactor pulse. In order to understand these photoneutron effects on the reactor kinetics, the KIFLE transient reactor-analysis code was modified to include the photoneutron groups associated with the beryllium. The time-dependent behavior of the reactor power was analyzed for small and large pulses, assuming several initial conditions including following several pulses during the day, and following a long steady-state power run. The results indicate that, for these types of initial conditions, the photoneutron contribution to the reactor pulse energy can have a few to tens of percent effect.
Date: June 1, 2009
Creator: Parma, Edward J., Jr.
Partner: UNT Libraries Government Documents Department

Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

Description: The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.
Date: October 1, 1978
Creator: Philbin, J.S.; Nelson, W.E. & Rosenstroch, B.
Partner: UNT Libraries Government Documents Department

Treatment of measurement uncertainties at the power burst facility

Description: The treatment of measurement uncertainty at the Power Burst Facility provides a means of improving data integrity as well as meeting standard practice reporting requirements. This is accomplished by performing the uncertainty analysis in two parts, test independent uncertainty analysis and test dependent uncertainty analysis. The test independent uncertainty analysis is performed on instrumentation used repeatedly from one test to the next, and does not have to be repeated for each test except for improved or new types of instruments. A test dependent uncertainty analysis is performed on each test based on the test independent uncertainties modified as required by test specifications, experiment fixture design, and historical performance of instruments on similar tests. The methodology for performing uncertainty analysis based on the National Bureau of Standards method is reviewed with examples applied to nuclear instrumentation.
Date: January 1, 1980
Creator: Meyer, L.C.
Partner: UNT Libraries Government Documents Department

A Numerical Model for Coupling of Neutron Diffusion and Thermomechanics in Fast Burst Reactors

Description: We develop a numerical model for coupling of neutron diffusion adn termomechanics in order to stimulate transient behavior of a fast burst reactor. The problem involves solving a set of non-linear different equations which approximate neutron diffusion, temperature change, and material behavior. With this equation set we will model the transition from a supercritical to subcritical state and possible mechanical vibration.
Date: November 1, 2008
Creator: Kadioglu, Samet Y.; Knoll, Dana A. & Oliveira, Cassiano De
Partner: UNT Libraries Government Documents Department

Godiva and Juliet Diagnostics CED-1 (IER-176)

Description: A suite of diagnostics are being proposed for use in the Juliet experiment (IER-128). In order to calibrate and test the diagnostics prior to use, the LLNL calibration facility and Godiva pulsed reactor will be used to provide intense sources of neutrons and gammas. Due to the similarities of the Godiva and Juliet radiation fields, the diagnostics being developed and tested for Juliet can also play an on-going role in diagnostics for Godiva as well as, perhaps, other critical assembly experiments. Similar work is also being conducted for IER-147 for the purpose of characterizing the Godiva radiation field in support of an upcoming international nuclear accident dosimetry exercise. Diagnostics developed and fielded under IER-147 can provide valuable data with respect to the neutron and gamma energy spectrums in the vicinity of Godiva which is relevant to the calibration of Juliet diagnostics.
Date: December 21, 2011
Creator: Scorby, J C
Partner: UNT Libraries Government Documents Department

Accident analysis for US fast burst reactors

Description: In the US fast burst reactor (FBR) community there has been increasing emphasis and scrutiny on safety analysis and understanding of possible accident scenarios. This paper summarizes recent work in these areas that is going on at the different US FBR sites. At this time, all of the FBR facilities have or in the process of updating and refining their accident analyses. This effort is driven by two objectives: to obtain a more realistic scenario for emergency response procedures and contingency plans, and to determine compliance with changing regulatory standards.
Date: September 1, 1994
Creator: Paternoster, R.; Flanders, M. & Kazi, H.
Partner: UNT Libraries Government Documents Department

Design guide for category V reactors transient reactors

Description: The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category V reactor structures, components, and systems.
Date: March 1, 1979
Creator: Brynda, W J; Karol, R C; Lobner, P R; Powell, R W & Straker, E A
Partner: UNT Libraries Government Documents Department

SPR-III 7-pin in-core fuel motion detection feasibility experiments

Description: Experiments were performed in the Sandia Pulsed Reactor to demonstrate the feasibility of in-core fuel motion detection. In these experiments fuel motion was mechanically induced and detected by 22 gamma sensitive detectors. A theoretical model using nonlinear perturbation theory was developed to determine the sensitivity of the detectors to fuel motion. This model, together with the measured detector responses, was unfolded using a least squares solution technique to determine the fuel motion. The unfolded motions were then compared with the actual motion to determine the accuracy of the technique. The fuel-mass resolution was +-5 to 10 g for fuel motions less than 75 g. For fuel motions greater than 150 g the resolutions were +-16 to 22 g. Thus, depending on the fuel-mass resolution required, it appears that the technique of in-core fuel motion detection is feasible for many reactor safety experiments.
Date: January 1, 1979
Creator: Wright, S.A.; McDaniel, P.J.; Powell, J.E. & Scott, W.H. Jr.
Partner: UNT Libraries Government Documents Department

Results of the initial test program for the Sandia Pulsed Reactor III (SPR III)

Description: This document presents a detailed discussion of the reactor including the mechanical and nuclear design characteristics. Also presented are the complete results of the Initial Approach to Critical and the Zero-and-Low Power testing programs. Reactivity worth measurements are given for such parameters as control element integral worth, Safety Block integral worth, and various materials (polyethylene, copper, lead, etc) as a function of position relative to the core. Subcritical reactivity measurements made during the approach to critical generally proved to be in reasonably good agreement with design values due to the good source-fuel-detector geometry possible with a reactor of this type. Subsequent dynamic measurements for reactivity worths are shown to be in good agreement with calculated results.
Date: August 1, 1976
Creator: Estes, B. F. & Reuscher, J. A.
Partner: UNT Libraries Government Documents Department


Description: H-Canyon Engineering (HCE) is evaluating the feasibility of processing material from the Super Kukla Prompt Burst Reactor, which operated at the Nevada Test Site from 1964 to 1978. This material is comprised of 90 wt % uranium (U) (at approximately 20% 235U enrichment) alloyed with 10 wt % molybdenum (Mo). The objective is to dissolve the material in nitric acid (HNO{sub 3}) in the H-Canyon dissolvers and then to process the dissolved material through H-Canyon First and Second Cycle solvent extraction. The U product from Second Cycle will be sent to the highly enriched uranium (HEU) blend down program. In the blend down program, enriched U from the 1EU product stream will be blended with natural U at a ratio of 1 part enriched U per 3.5 parts natural U to meet a reactor fuel specification of 4.95% 235U before being shipped for use by the Tennessee Valley Authority (TVA) in its nuclear plants. The TVA specification calls for <200 mg Mo/g U (200 ppm). Since natural U has about 10 mg Mo/g U, the required purity of the 1EU product prior to blending is about 800 mg Mo/g U, allowing for uncertainties. HCE requested that the Savannah River National Laboratory (SRNL) define a flowsheet for the safe and efficient processing of the U-10Mo material. This report presents a computational model of the solvent extraction portion of the proposed flowsheet. The two main objectives of the computational model are to demonstrate that the Mo impurity requirement can be met and to show that the solvent feed rates in the proposed flowsheet, in particular to 1A and 1D Banks, are adequate to prevent refluxing of U and thereby ensure nuclear criticality safety. SASSE (Spreadsheet Algorithm for Stagewise Solvent Extraction), a Microsoft Excel spreadsheet that supports Argonne National Laboratory's proprietary AMUSE (Argonne ...
Date: May 31, 2007
Creator: Laurinat, J
Partner: UNT Libraries Government Documents Department

Analysis and Numerical Solution for Multi-Physics Coupling of Neutron Diffusion and Thermomechanics in Spherical Fast Burst Reactors

Description: Coupling neutronics to thermomechanics is important for the analysis of fast burst reactors, because the criticality and safety study of fast burst reactors heavily depends on the thermomechanical behavior of fuel materials. For instance, the shut down mechanism or the transition between super and sub-critical states are driven by the fuel material expansion or contraction. The material expansion or contraction is due to temperature gradient which results from fission power. In this paper, we introduce a numerical model for coupling of neutron diffusion and thermomechanics in fast burst reactors. We also provide some analysis of the coupled system. We studied material behaviors corresponding to different levels of power pulses.
Date: May 1, 2009
Creator: Kadioglu, Samet Y.; Knoll, Dana A. & Oliveira, Cassiano de
Partner: UNT Libraries Government Documents Department

Final report on LDRD project 105967 : exploring the increase in GaAs photodiode responsivity with increased neutron fluence.

Description: A previous LDRD studying radiation hardened optoelectronic components for space-based applications led to the result that increased neutron irradiation from a fast-burst reactor caused increased responsivity in GaAs photodiodes up to a total fluence of 4.4 x 10{sup 13} neutrons/cm{sup 2} (1 MeV Eq., Si). The silicon photodiodes experienced significant degradation. Scientific literature shows that neutrons can both cause defects as well as potentially remove defects in an annealing-like process in GaAs. Though there has been some modeling that suggests how fabrication and radiation-induced defects can migrate to surfaces and interfaces in GaAs and lead to an ordering effect, it is important to consider how these processes affect the performance of devices, such as the basic GaAs p-i-n photodiode. In this LDRD, we manufactured GaAs photodiodes at the MESA facility, irradiated them with electrons and neutrons at the White Sands Missile Range Linac and Fast Burst Reactor, and performed measurements to show the effect of irradiation on dark current, responsivity and high-speed bandwidth.
Date: January 1, 2008
Creator: Blansett, Ethan L.; Geib, Kent Martin; Cich, Michael Joseph; Wrobel, Theodore Frank; Peake, Gregory Merwin; Fleming, Robert M. et al.
Partner: UNT Libraries Government Documents Department

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

Description: The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.
Date: August 1, 2009
Creator: Rosenthal, Murray Wilford
Partner: UNT Libraries Government Documents Department

Recent operational history of the new Sandia Pulsed Reactor III (SPR III)

Description: The Sandia Pulsed Reactor III (SPR III) is a fast-pulse research reactor which was designed and built at Sandia Laboratories and achieved criticality in August 1975. The reactor is now characterized and is in an operational configuration. The core consists of 18 fuel plates (258 kg fuel mass) of fully enriched uranium alloyed with 10 wt.% molybdenum. It is arranged in an annular configuration with an inside diameter of 17.78 cm, an outside diameter of 29.72 cm, and a height of 35.9 cm. The reactor core uses reflectors of copper and aluminum for control and an external bolting arrangement to secure the fuel plates. SPR III and SPR II are operated on an interchangeable basis using the same facility and control system. As of June 1977, SPR III has had over 240 operations with core temperatures up to 541/sup 0/C.
Date: January 1, 1977
Creator: Schmidt, T. R.; Estes, B. F. & Reuscher, J. A.
Partner: UNT Libraries Government Documents Department