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MUSE-4 experiment measurements and analysis.

Description: This report presents a review of the activities performed by the five teams involved in the MUSE-4 experimental program. More details are provided on the contribution by ANL during the year 9/02 to 9/03. The ANL activity consisted both in direct participation in the experimental measurements and in the physics analysis of the experimental data, mainly for the reactivity level, adjoint flux and fission rate distributions and the analysis of dynamic measurements for reactivity determination techniques in subcritical systems. The results provided to complete the Benchmark organized by the OECD and the CEA on the experiment MUSE-4 are also presented. Deterministic calculations have been performed via the ERANOS code system in connection with JEF2.2, ENDF/B-V and ENDF/B-VI data files.
Date: January 13, 2004
Creator: Aliberti, G.; Imel, G. & Palmiotti, G.
Partner: UNT Libraries Government Documents Department

Mark 22 Reactivity

Description: Calculations for reactivity held in control rods have underpredicted the observed Mark 22 reactivity. Reactivity predictions by charge designers have accounted for this by including large biases which change with exposure and reactor region. The purpose of this study was to thoroughly investigate the methods and data used in the reactivity calculations. The goal was to identify errors and improvements and make necessary corrections.
Date: July 2, 2001
Creator: Buckner, M.R.
Partner: UNT Libraries Government Documents Department

Variational reactivity estimates: new analyses and new results

Description: A modified form of the variational estimate of the reactivity worth ofa perturbation was previously developed to extend the range of applicability of variational perturbation theory for perturbations leading to negative reactivity worths. Recent numerical results challenged the assumptions behind the modified form. In this paper, more results are obtained, leading to the conclusion that sometimes the modified form extends the range ofapplicability of variational perturbation theory for positive reactivity worths as well, and sometimes the standard variational form is more accurate for negative-reactivity perturbations. In addition, this paper proves that using the exact generalized adjoint function would lead to an inaccurate variational reactivity estimate when the error in the first-order estimate is large; the standard generalized adjoint function, an approximation to the exact one, leads to Lore accurate results. This conclusion is also demonstrated numerically. Transport calculations use the PARTISN multi group discrete ordinates code
Date: January 1, 2009
Creator: Favorite, Jeffrey A
Partner: UNT Libraries Government Documents Department

Criticality measurements for SNM accountability

Description: Based on extensive operating experience with the Godiva IV fast metal burst assembly at Los Alamos National Laboratory, the authors were able to create data plots for reactivity worths of standard configurations at various temperatures and room return locations. These plots show that the material uncertainties in criticality measurements are within {+-} 20 grams out of the 65.4 kilogram HEU Godiva core. This is superior to active neutron well coincidence counter (AWCC) measurements. The criticality measurements have the additional advantage of not requiring disassembly of the reactor. No disassembly means the measurement takes less time--it can be done during each operation--and there is less dose to measurement personnel.
Date: March 1, 1998
Creator: Bohman, J.; Martin, E.R.; Butterfield, K. & Paternoster, R.
Partner: UNT Libraries Government Documents Department

Test Pile Reactivity Loss Due to Trichloroethylene

Description: The presence of trichloroethylene in the test pile caused a continual decrease in pile reactivity. A system which removed, purified, and returned 12,000 cfh helium to the pile has held contamination to a negligible level and has permitted normal pile operation.
Date: March 9, 2001
Creator: Plumlee, K. E.
Partner: UNT Libraries Government Documents Department

Effect of reflector location at array criticality

Description: The motion of a thick reflector away from a critical array of subcritical units of fissile material results in an array reactivity loss. The fraction of the total reactivity worth of a reflector is related to the distance a reflector is located from the surfaces of an array. The magnitude of reactivity associated with the reflector location may be used in the assessment of nuclear criticality safety of operations with fissile materials to establish margins of subcriticality. An equivalence is established between the reactivity associated with reflector location and the reactivity associated with the thickness of a water reflector closely fitting an array.
Date: November 1, 1980
Creator: Thomas, J.T.
Partner: UNT Libraries Government Documents Department


Description: In this paper the effect of changing from the traditional NTP coolant, hydrogen, to several alternative coolants were studied. Hydrogen is generally chosen as an NTP coolant, since its use maximizes the specific impulse for a given operating temperature. However, there are situations in which it may not be available or optimal. The alternative coolants which were considered are ammonia, methane, and carbon dioxide. A particle bed reactor (PBR) generating 200 MW and cooled by hydrogen was used as the baseline against which all the comparisons were made. Both 19 and 37 element cores were considered. The larger number of elements was found to be necessary in the case of carbon dioxide. The coolant reactivity worth was found to be directly proportional to the hydrogen coolant content. It was found that due to differences in the thermophysical proportions of the coolant that it would not be possible to use one reactor for all the coolants. The reactor would have to be constructed specifically for a coolant type.
Date: January 12, 1992
Creator: Selcow, E. C.; Davis, R.; Perkins, K.; Ludewig, H. & Cerbone, R.
Partner: UNT Libraries Government Documents Department

Criticality calculations for the VR-1 reactor with IRT-3M-HEU fuel and IRT-4MLEU fuel.

Description: At The request of the Czech Technical University in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. For core configurations C1 and C2, criticality calculations were done for cases with all control rods at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were done for the C1 core configuration. Finally the reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations.
Date: January 17, 2007
Creator: Hanan, N. A. & Matos, J. E.
Partner: UNT Libraries Government Documents Department

Tradeoff of sodium void worth and burnup reactivity swing: Impacts on balance safety position in metallic-fueled cores

Description: A study has been conducted to investigate the effect of a lower sodium void worth on the consequences of severe accidents in metallic-fueled sodium-cooled reactors. Four 900 MWth designs were used for the study, where all of the reactor cores were designed based on the metallic fuel of the Integral Fast Reactor (IFR) concept. The four core designs each have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation was to determine the differences in severe accident response for the four core designs, in order to estimate the improvement in overall safety that could be achieved from a reduction in the sodium void worth for reactor cores which use a metallic fuel form.
Date: October 1, 1994
Creator: Wigeland, R. A.; Turski, R. B. & Pizzica, P. A.
Partner: UNT Libraries Government Documents Department

Integral capture cross-section measurements in the CFRMF for LMFBR control materials

Description: Integral capture-cross sections for separated isotopes of Eu and Ta are reported for measurements in the Coupled Fast Reactivity Measurements Facility (CFRMF). These cross sections along with that measured in the CFRMF for $sup 10$B(n,$alpha$) provide an absolute standard for evaluating the relative reactivity worth of Eu$sub 2$O$sub 3$, B$sub 4$C and Ta in neutron fields typical of an LMFBR core. Based on these measurements and for neutron fields characterized by the $sup 235$U:$sup 238$U reaction rate spectral index ranging from 23 to 50, the infinitely dilute relative worth of Eu$sub 2$O$sub 3$ has been estimated to be 25 to 40 percent higher than that for B$sub 4$C and 80 percent to 100 percent higher than that for Ta. 11 references. (auth)
Date: January 1, 1975
Creator: Anderl, R.A.; Harker, Y.D.; Turk, E.H.; Nisle, R.G. & Berreth, J.R.
Partner: UNT Libraries Government Documents Department

Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

Description: Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron. Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for ...
Date: April 17, 2002
Creator: Renier, J.A.
Partner: UNT Libraries Government Documents Department

Measurements of delayed neutron parameters for U-235 and Np-237

Description: Delayed neutrons are emitted by excited nuclei formed in beta decay of fission products called delayed neutron precursors. About 1% of the total neutrons released in fission are delayed neutrons; however, this small fraction plays an important role in nuclear reactor control. The delayed neutrons determine the time-dependent behavior of reactors, and knowledge of parameters used to predict neutron emission rate is essential for establishing reactivity worths. The delayed neutron yields, decay constants, and the absolute yield for the six-group delayed neutrons have been measured for U-235 and Np-237. This experiment has been called for in the forecast of experiments needed to support operations in the US. The bare U-235 metal assembly Godiva IV at the Los Alamos Critical Experiment Facility (LACEF) provided the source of neutrons. Godiva IV generated about 10{sup 7} total fissions in the samples for the infinite and instantaneous irradiation needed to accentuate the shorter and longer-lived groups of delayed neutrons. The detection system used in the experiment consisted of 20 He-3 tubes embedded in a polyethylene cylinder. The delayed neutron activity resulting from the fast neutron-induced fission has been measured. The measured absolute yield for U-235 was determined to be 0.0163 {+-} 0.009 neutrons/fission. This value compares very well with the well-established Keepin absolute yield of 0.0165 {+-} 0.0005. The newly measured absolute yield value for Np-237 was 0.0126 {+-} 0.0007, which compares well to the recently reported value of 0.0129 {+-} 0.0004 by Saleh and Parish. The measured values for U-235 are corroborated with period (e-folding time) versus reactivity calculations.
Date: July 1997
Creator: Loaiza, D.
Partner: UNT Libraries Government Documents Department

The central void reactivity in the Oak Ridge enriched uranium (93.2) metal sphere

Description: The central reactivity void worth was measured in the Oak Ridge unmoderated and unreflected uranium (93.20 wt% {sup 235}U) metal sphere by replacement measurements in a small (0.460-cm-diam) central spherical region in an 8.7427-cm-radius sphere. The central void worth was 9.165 {+-} 0.023 cents using the delayed neutron relative abundances and decay constants of Keepin, Wimett, and Zeigler to obtain the reactivity in cents from the stable reactor period measurements using the Inhour equation. This value is slightly larger than measurements with GODIVA 1 with larger cylindrical samples of uranium (93.70 wt% {sup 235}U) in the center: 135.50 {+-} 0.12 cents/mole for GODIVA 1 and 138.05 {+-} 0.34 cents/mole for the Oak Ridge sphere measurements, and the difference could be due to sample size effect. The central worth in {Delta}k units was calculated by neutron transport theory methods to be 6.02 {+-} 0.01 x 10{sup {minus}4} {Delta}k. The measured and calculated values are related by the effective delayed neutron fraction. The value of the effective delayed neutron fraction obtained in this way from the Oak Ridge sphere is 0.00657 {+-} 0.00002, which is in excellent agreement with that obtained from GODIVA 1 measurements, where the effective delayed neutron fraction was determined as the increment between delayed and prompt criticality and was 0.0066. From these Oak Ridge measurements, using the delayed neutron parameters of ENDF-B/VI to obtain the reactivity from the stable reactor period measurements, the central void worth is 7.984 {+-} 0.021 cents, and the inferred effective delayed neutron fraction is 0.00754. This central void worth and effective delayed neutron fractions are 14.2% higher than those obtained from use of the Keepin et al. delayed neutron data and produce a value of delayed neutron fraction in disagreement with GODIVA 1 measurements, thus questioning the usefulness of the relative abundances and ...
Date: March 1, 1997
Creator: Milhalczo, J.T.; Lynn, J.J. & Taylor, J.R.
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

Central worth and spectral measurements in the GCFR. Phase I assembly

Description: Central fission and capture rates, the central neutron spectrum and the reactivity worths of small samples were measured at the core center of the GCFR Phase I Assembly, the initial benchmark GCFR mockup assembly. Results of these measurements and comparisons with calculations are reported. (auth)
Date: November 1, 1975
Creator: Morman, J.A.; Bhattacharyya, S.K.; Smith, D.M.; McKnight, R.D.; Yule, T.J. & Bohn, E.M.
Partner: UNT Libraries Government Documents Department

Generalized perturbation theory using two-dimensional, discrete ordinates transport theory

Description: Perturbation theory for changes in linear and bilinear functionals of the forward and adjoint fluxes in a critical reactor has been implemented using two-dimensional discrete ordinates transport theory. The computer program DOT IV was modified to calculate the generalized functions GAMMA and GAMMA*. Demonstration calculations were performed for changes in a reaction-rate ratio and a reactivity worth caused by system perturbations. The perturbation theory predictions agreed with direct calculations to within about 2%. A method has been developed for calculating higher lambda eigenvalues and eigenfunctions using techniques similar to those developed for generalized functions. Demonstration calculations have been performed to obtain these eigenfunctions.
Date: June 1, 1980
Creator: Childs, R.L.
Partner: UNT Libraries Government Documents Department

Critical experiments in support of the CNPS (Compact Nuclear Power Source) program

Description: Zero-power static and kinetic measurements have been made on a mock-up of the Compact Nuclear Power Source (CNPS), a graphite moderated, graphite reflected, U(19.9% /sup 235/U) fueled reactor design. Critical configurations were tracked from a first clean configuration (184 most central fuel channels filled and all control rod and heat pipe channels empty) to a fully loaded configuration (all 492 fuel channels filled, core-length stainless steel pipe in the twelve heat-pipe channels, and approximately half-core-length boron carbide in the outer 4 control rod channels. Reactor physics data such as material worths and neutron lifetime are presented only for the clean and fully loaded configurations.
Date: January 1, 1988
Creator: Hansen, G.E.; Audas, J.H.; Martin, E.R.; Pederson, R.A.; Spriggs, G.D. & White, R.H.
Partner: UNT Libraries Government Documents Department

Reckoning THOR

Description: Theoretical computation of the Los Alamos National Laboratory's critical assembly THOR (a thorium-reflected plutonium sphere) yields a high eigenvalue when compared to the experimentally measured eigenvalue. Several calculational improvements are investigated in an effort to reduce the discrepancy. Finally, the experimental procedure of reducing the raw configuration to clean specifications is reviewed.
Date: May 1, 1981
Creator: Kidman, R.B.
Partner: UNT Libraries Government Documents Department

Calculation of RABBIT and Simulator Worth in the HFIR Hydraulic Tube and Comparison with Measured Values

Description: To aid in the determinations of reactivity worths for target materials in a proposed High Flux Isotope Reactor (HFIR) target configuration containing two additional hydraulic tubes, the worths of cadmium rabbits within the current hydraulic tube were calculated using a reference model of the HFIR and the MCNP5 computer code. The worths were compared to measured worths for both static and ejection experiments. After accounting for uncertainties in the calculations and the measurements, excellent agreement between the two was obtained. Computational and measurement limitations indicate that accurate estimation of worth is only possible when the worth exceeds 10 cents. Results indicate that MCNP5 and the reactor model can be used to predict reactivity worths of various samples when the expected perturbations are greater than 10 cents. The level of agreement between calculation and experiment indicates that the accuracy of such predictions would be dependent solely on the quality of the nuclear data for the materials to be irradiated. Transients that are approximated by ''piecewise static'' computational models should likewise have an accuracy that is dependent solely on the quality of the nuclear data.
Date: September 8, 2005
Creator: Slater, C. O.
Partner: UNT Libraries Government Documents Department