Annular Core Research Reactor Research Articles

Monte Carlo Perturbation Analysis of Fuel Temperature Variations in the MCNP Model of the Annular Core Research Reactor

The Annular Core Research Reactor (ACRR) Monte Carlo N-Particle (MCNP) model is used by ACRR reactor operators and experiment designers at Sandia National Laboratories for a variety of computational calculations ranging from reactor kinetics parameter estimates and safety analyses to experimental planning. To understand the dominant source of uncertainty within the MCNP model, perturbations in temperature were applied to individual ACRR MCNP fuel rods. Fuel rod temperatures were randomly sampled from a uniform distribution from operational temperatures to quantify temperature-related uncertainty effects. Stochastic mixing was used to blend the cross sections of the desired temperatures using the MCNP continuous and Thermal Neutron Scattering Treatment [S(α,β)] libraries in ENDF/B-VII.1. This uncertainty analysis produced a 640 row × 640 column correlation and covariance matrix of the neutron energy spectra. Positive covariance was produced around the 1-MeV region and the 0.2-eV region. Correlation was found in the thermal and fast energy regions, but no correlation was observed in the slowing-down energy region because interactions in this region are not dominated by fuel.

Method for calculating delayed gamma-ray response in the ACRR Central Cavity using MCNP

This document presents the process for a new method developed for the characterization of the delayed gamma-ray radiation fields in pulse reactors like the Annular Core Research Reactor (ACRR). The environments used to test this method in the ACRR Central Cavity were various standard irradiation configurations that used various amounts of moderators and shielding. All environment configurations used the same fission product gamma-ray source energy spectrum. This method required the use of fission sites recorded in the Monte Carlo N-Particle (MCNP) KCODE source tape. A FORTRAN script was written to translate and extract the coordinates for the fission sites. The fission sites were then input into an MCNP source mode input file. Using a MATLAB script, a parametric analysis was performed and determined that 10K fission sites are an appropriate number of coordinates to converge to the correct answer. The method gave excellent results as compared to previous methods. This method can be applied to other pulse research reactors as well.

Razorback – A reactor transient analysis code for large rapid reactivity additions in a natural circulation research reactor

Research reactors play an important role in higher education, scientific research, and medical radioisotope production around the world. It is thus important to ensure the safety of facility workers and the public. This work presents a new reactor transient analysis code referred to as Razorback. The code has been developed for the evaluation of large rapid reactivity addition in research reactors. Its initial focus is the Annular Core Research Reactor (ACRR) at Sandia National Laboratories. Results have been validated against ACRR pulse operations. Razorback models the reactor kinetics, fuel element heat transfer, fuel element thermal expansion, and natural circulation coolant channel thermal-hydraulic response. Simulation results for ACRR pulse operations are shown to agree very well with operational data obtained from the ACRR. Razorback is expected to be a valuable tool for ACRR pulse performance prediction and ACRR reactor safety analyses.

Boron filled siloxane polymers for radiation shielding

The purpose of the present work was to evaluate changes to structure-property relationships of 10B filled siloxane-based polymers when exposed to nuclear reactor radiation. Highly filled polysiloxanes were synthesized with the intent of fabricating materials that could shield high neutron fluences. The newly formulated materials consisted of cross-linked poly-diphenyl-methylsiloxane filled with natural boron and carbon nanofibers. This polymer was chosen because of its good thermal and chemical stabilities, as well as resistance to ionizing radiation thanks to the presence of aromatic groups in the siloxane backbone. Highly isotopically enriched 10B filler was used to provide an efficient neutron radiation shield, and carbon nanofibers were added to improve mechanical strength. This novel polymeric material was exposed in the Annular Core Research Reactor (ACRR) at Sandia National Labs to five different neutron/gamma fluxes consisting of very high neutron fluences within very short time periods. Thermocouples placed on the specimens recorded in-situ temperature changes during radiation exposure, which agreed well with those obtained from our MCNP simulations. Changes in the microstructural, thermal, chemical, and mechanical properties were evaluated by SEM, DSC, TGA, FT-IR NMR, solvent swelling, and uniaxial compressive load measurements. Our results demonstrate that these newly formulated materials are well-suitable to be used in applications that require exposure to different types of ionizing conditions that take place simultaneously.

Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR) is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

Characterization of Novel Calorimeters in the Annular Core Research Reactor *

A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR) to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field – a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP) bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ) contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified) cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ) cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response.

Modeling, Calibration, and Verification of a Fission Chamber for ACRR Experimenters

When performing research at a reactor facility, experimenters often need to determine the neutron fluence achieved during an operation. Facilities typically provide guidance in the form of neutron fluence per megajoule (MJ) or through passive dosimetry results. After experiment completion, there is sometimes a delay of several days (or weeks) before the passive dosimetry results are available. In the interim, an experimenter does not have confirmation that the desired irradiation levels were reached. Active dosimetry may provide an estimate of neutron fluxes, but few active detectors are available that have been calibrated to measure neutron fluxes obtained inside the Annular Core Research Reactor (ACRR) central cavity environment. For past experiments at the ACRR, the neutron fluence was calculated by integrating the response of a fission chamber rate detection signal and then normalizing this integral to fluence determined from passive dosimetry. An alternative method of directly measuring neutron flux is desired; the new methodology described provides a complete neutron flux profile after a reactor pulse, utilizing fission chamber physics in combination with a compensating ion chamber to extract and convert a current signal to neutron flux as a function of time.

Test simulation of neutron damage to electronic components using accelerator facilities

The purpose of this work is to demonstrate equivalent bipolar transistor damage response to neutrons and silicon ions. We report on irradiation tests performed at the White Sands Missile Range Fast Burst Reactor, the Sandia National Laboratories (SNL) Annular Core Research Reactor, the SNL SPHINX accelerator, and the SNL Ion Beam Laboratory using commercial silicon npn bipolar junction transistors (BJTs) and III–V Npn heterojunction bipolar transistors (HBTs). Late time and early time gain metrics as well as defect spectra measurements are reported.

Development of an Active Detector for the Characterization of the Late-Time Radiation Environment from a Reactor Pulse

Abstract This paper discusses the use of a commercially available 235U fission chamber, with a matching compensating ion chamber, originally sold as a single-ended detector with the signal conducted over the shield of a coaxial cable. The authors designed an aluminum housing that isolates the two detectors and converts the signals to full differential mode as a noise-reduction technique. The signals are processed using the switched resistor technique to extend the signal range to longer times from the peak of the pulse [Luker, S. M., Griffin, P. J., King, D. B., and Suo-Anttila, A. J., “Improved Diagnostics for Analysis of a Reactor Pulse Radiation Environment,” 13th International Symposium on Reactor Dosimetry, Akersloot, Netherlands, May 25, 2008, pp. 4–6.]. The newly configured fission chamber assembly has been used at the annular core research reactor at Sandia National Laboratories to provide a high-fidelity characterization of the neutron time profile from a pulsed operation.

EPR/PTFE Dosimetry for Test Reactor Environments

Abstract In 1992 the Radiation Metrology Laboratory (RML) at Sandia National Laboratories implemented EPR/Alanine capabilities for use in routine and calibration activities at its Co-60 and pulsed-power facilities. At that time it also investigated the usefulness of the system for measurement of absorbed dose in the mixed neutron/photon environments of reactors such as the Sandia Pulsed Reactor and the Annular Core Research Reactor used for hardness testing of electronics. The RML concluded that the neutron response of alanine was a sufficiently high fraction of the overall dosimeter response that the resulting uncertainties in the photon dose would be unacceptably large for silicon-device testing. However, it also suggested that non-hydrogenous materials such as polytetrafluoroethylene (PTFE) would exhibit smaller neutron response and might be useful in mixed environments. Preliminary research with PTFE in photon environments indicated considerable promise, but further development was not pursued at that time. Because of renewed interest in absorbed dose measurements that could better define the individual contributions of photon and neutron components to the overall dose delivered to a test object, the RML has re-initiated the development of an EPR/PTFE dosimetry system. This paper presents a summary of the research, a description of the EPR/PTFE dosimetry system, and recommendations for preparation and fielding of the dosimetry in photon and mixed neutron/photon environments.

Reactor Pulse-Repeatability Studies at the Annular Core Research Reactor

Abstract The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO2–BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the 9Be(γ, n)8Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times.

An Approach to Determining the Uncertainty in Reactor Test Objects Using Deterministic and Monte Carlo Methods

Abstract A procedure for determining an estimate of the uncertainty in test objects components using discrete ordinates and Monte Carlo methods has been developed. The procedure is described, and a review of the advantages and disadvantages of these methods is presented with the emphasis on the advantage of employing both transport methods in analysis. The discrete ordinates-based techniques use the DOORS and DANTSYS radiation transport codes (with the SUSD3D code for the determination of sensitivity of the response parameter to cross sections), and the Monte Carlo-based techniques primarily employ the MCNP code (using the general MCNP perturbation operator in a manner that determines the sensitivities to transport cross sections). The focus of the analysis presented is on experiments that are preformed at the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL), and includes uncertainty contributed by the uncertainties in the source spectrum, in response functions, and in transport cross sections.

Benchmark Experiments/Calculations of Neutron Environments in the Annular Core Research Reactor

Abstract Benchmark experiments using spherical test objects were performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor. The experiments were performed with 10.16 cm and 17.78 cm diameter aluminum (AI6061) and high-density polyethylene spheres that were essentially solid with cavities scrolled out along the equator to allow the insertion of activation foils and/or sulfur pellets. The neutron monitor foils were selected to cover a wide range of reaction energies. The reactor environments were modeled in detail using Monte Carlo N-Particle eXtended (mcnpx). The experimental results were compared to the Monte Carlo calculations of the reaction rates of each of the foils at various depths in the spheres produced by mcnpx. The comparison includes a complete treatment of the uncertainties.

Thermal radiation in gas core nuclear reactors for space propulsion

A diffusive model of the radial transport of thermal radiation out of a cylindrical core of fissioning plasma is presented. The diffusion approximation is appropriate because the opacity of uranium is very high at the temperatures of interest (greater than 3000 K). We make one additional simplification of assuming constant opacity throughout the fuel. This allows the complete set of solutions to be expressed as a single function. This function is approximated analytically to facilitate parametric studies of the performance of a test module of the nuclear light bulb gas-core nuclear-rocket-engine concept, in the Annular Core Research Reactor at Sandia National Laboratories. Our findings indicate that radiation temperatures in range of 4000-6000 K are attainable, which is sufficient to test the high specific impulse potential (approximately 2000 s) of this concept. 15 refs.

DF-4 Analysis Using SCDAP/RELAP5

In this paper the SCDAP/RELAPS severe accident analysis computer code, developed at the Idaho National Engineering Laboratory, is used to analyze the fourth in a series of debris formation experiments. The debris formation-four (DF-4) experiment deals with heatup and meltdown of a boiling water reactor (BWR)-representative fuel and control blade assembly segment, performed in the Annular Core Research Reactor at Sandia National Laboratories. The DF-4 experiment provides data that are used to validate core damage progression and BWR-specific models to gain an understanding of the phenomena occurring in the bundle during a severe BWR accident and to identify additional modeling needed in severe accident codes. The SCDAP/RELAPS model used for this analysis accurately predicts the key damage events, which include control blade melting, channel box relocation and runaway oxidation, the order and timing of these events, and the maximum bundle temperature. From these analytical calculations, an accident scenario and insights into phenomena occurring during a severe BWR accident are developed.

Considerations in the Design and Implementation of Control Laws for the Digital Operation of Research Reactors

In this paper factors relevant to the design and implementation of digital controllers for research reactors are discussed with emphasis on the rationale for incorporating a system model in the control law. For this purpose, proportional-integral-derivative and period-generated control are compared. The latter is a model-based technique that achieves excellent trajectory tracking of nonlinear systems. It does this by combining feedback and feedforward control action in a manner that cancels the effects of the system's dynamics on the controller's performance. Model-based control is also superior in that it permits replication of some of the functions that humans perform when exercising control. In particular, models can be used to predict expected plant response and thereby facilitate diagnosis. Experimental results obtained from trials of digital controllers on both the 5-MW (thermal) Massachusetts Institute of Technology Research Reactor and the Annular Core Research Reactor that is operated by Sandia National Laboratories are included.

Measurements of the Total Pressure from Irradiated (U,Pu)-Mixed Oxide

Irradiated (U.Pu)-mixed oxide with 5% burnup was heated in the pulsed Annular Core Research Reactor at Sandia National Laboratories. The tests were typical of prompt Bethe-Tait excursions in terms of heating method (nuclear fission), heating period (milliseconds), and temperatures attained (up to 7700 K).Fission products provided high pressures at temperatures at which fresh fuel shows only a negligible vapor pressure. Fission product release became measurable as soon as the temperature exceeded the steady-state irradiation temperature of the fuel sample. The fission product pressures reached 1.3 to 2.5 MPa at 3000 K over solid fuel, and 2.5 to 5 MPa at 4000 K over liquid fuel. The total amount of fission product released corresponded to ∼30 to 75% of the fission gas inventory. The amount of fission product released increased with the fuel heating rate.Under rapid heating, the total pressure over irradiated (U,Pu) oxide is controlled by a suppression mechanism. At any given temperature, the gaseous components (xenon, cesium, and ambient gas) suppress fuel boiling if their pressure pgas is higher than the fresh fuel saturation vapor pressure psat of unirradiated fuel. If psat exceeds pgas, the total pressure is, to a first approximation, equal to Psat. Under millisecond heating, the total pressure from irradiated fuel may be taken as ptot = max(Pgas,Psat).

Simulation fidelity issues in reactor irradiation of electronics-reactor environments

Fast burst reactors, such as SPR III, and pool-type reactors, such as the University of Illinois TRIGA and the Sandia Annular Core Research Reactor, are used as sources of neutrons and/or gamma rays in transient radiation effects in electronics research and testing. The authors identify the important simulation fidelity issues, compare radiation environments produced in representative reactors, and describe custom environments that have been achieved to better meet experimenter requirements.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Contributions from the ACRR in-pile experiments to the understanding of key phenomena influencing unprotected loss of flow accident simulations in LMFBRs

This paper summarizes the recently terminated in-pile experiments in the Annular Core Research Reactor (ACRR) of the Sandia National Laboratories that have been performed to study the behavior of fast reactor fuel under different conditions during an unprotected loss of flow (ULOF) accident in an LMFBR. Three groups of experiments are covered. The FD + STAR tests that study the fuel disruption and initial fuel motion in voided regions, the TRAN experiments on flowing and freezing of molten fuel in structures, and the equation of state tests with fresh and irradiated fuel under prompt burst conditions. Conclusions are drawn on the contributions which the results from these tests make to the understanding and simulation of key phenomena under ULOF accident conditions in LMFBR's.