Total Number Of Fissions Research Articles

Improved Cumulative Fission Yield Measurements with Fission Spectrum Neutrons on 238U

Two isotopically-characterized targets of high-purity (99.961±0.002%) 238U metal and a neutron dosimetry package were exposed to a pulsed neutron irradiation using the Godiva IV critical assembly at the National Criticality Experiments Research Center to measure the integral fission product yields produced in a Watt fission neutron-energy spectrum. Following irradiation, the fissioned targets were counted with two broad energy germanium detectors. Estimates of the total fissions for each target were made using measurements of well-known fission products. The total number of fissions in each target were then used to measure the fission yields of 33 fission products, 18 of which are published as new measurements. Twelve of the 18 isotopes were in agreement with ENDF/B-VIII.0 using the zeta test and treating ENDF/B-VIII.0 as the true mean: 84mBr, 93Y, 94Y, 104Tc, 134I, 135I, 138Xe, 141Ba, 142Ba, 142La, 146Ce, and 149Nd. The ENDF/B-VIII.0 fission yield for 139Ba was only slightly outside of 1-σ relative to the uncertainty measured in this work. Uncertainties for 5 fission product yields were markedly improved over those reported in ENDF/B-VIII.0: 128Sn, 129Sb, 130gSb, 131mTe, and 133I. The fission yields measured for 128Sn, 131mTe, and 133I were in good agreement with the ENDF/B-VIII.0 predicted fission yields. However, fission yields measurements of the isotopes 129Sb, and 130gSb deviated from the ENDF/B-VIII.0 value. Both of these isotopes reside near the doubly magic nuclear shell closures of 50 protons and 82 neutrons. These fission yields may be suppressed because of the odd-odd nuclear structure of 130gSb and the odd-even structure of 129Sb and isomer splitting. Comparisons to other recent work and a brief review with citations are included to support this conclusion.

Metrology for Transient Reactor Characterization Using Uranium Wires

The Transient Reactor Test (TREAT) facility, located at Idaho National Laboratory, restarted transient operations in 2018 following an extended shutdown. It is of interest to establish a methodology and capability to obtain an accurate estimate of the total number of fissions produced in a fissionable test item during a transient at TREAT. Uranium wires were irradiated in TREAT as part of a transient prescription test program, and gamma-ray spectrometry was performed on the wires following irradiation using a high-purity germanium detector. Many fission products are useful for estimating the number of fissions produced in a sample using gamma-ray spectrometry; at TREAT with the time periods used for analysis, the isotopes of interest include 95Zr, 95Nb, 103Ru, 140Ba, and 140La. The number of fissions per gram of 235U determined from these measurements establishes an estimate for future experiments to be performed in the core when a similar configuration is used with a similar transient prescription.

Monte Carlo simulations and experimental results on neutron production in the uranium spallation target QUINTA irradiated with 660 MeV protons

The activation experiment was performed using the accelerated beam of Phasotron accelerator at the Joint Institute for Nuclear Research (JINR). The natural uranium spallation target QUINTA was irradiated with protons with energy 660 MeV. Monte Carlo simulations of neutron production were performed using the Geant4 code. The number of leakage neutrons from the sections of the uranium target surrounded by the lead shielding and the number of leakage neutrons from lead were determined. The total number of fissions in the setup QUINTA was determined. Experimental values of reaction rates for the produced nuclei in the 127I sample were obtained and several values of reaction rates were compared with the results of simulations. Experimentally determined fluence of neutrons in energy interval 10–175 MeV using the (n,xn) reactions in the 127I(NaI) sample was compared with the results of simulations. Possibility of transmutation of the long-lived radionuclide 129I in the QUINTA setup was estimated.

Calculating Time-Integral Quantities in Depletion Calculations

A method referred to as tally nuclides is presented for accurately and efficiently calculating the time-step averages and integrals of any quantities that are weighted sums of atomic densities with constant weights during the step. The method allows all such quantities to be calculated simultaneously as a part of a single depletion solution with existing depletion algorithms. Examples of results that can be extracted include step-average atomic densities and macroscopic reaction rates, the total number of fissions during the step, and the amount of energy released during the step. The method should be applicable with several depletion algorithms, and the integrals or averages should be calculated with an accuracy comparable to that reached by the selected algorithm for end-of-step atomic densities. The accuracy of the method is demonstrated in depletion calculations using the Chebyshev rational approximation method. As an example of a possible use, we demonstrate how the ability to calculate energy release in depletion calculations can be used to determine the accuracy of the normalization in a constant-power burnup calculation during the calculation without a need for a reference solution.

A point kinetics model of the Y12 accident

A point kinetics model of the 1958 Y12 criticality accident is presented and the relevant parameters are obtained from review of the literature and novel analysis. The results of the simulation are broadly consistent with the literature in terms of the fission rate as a function of time and the total number of fissions. The literature is inconsistent regarding whether the solution boiled during this accident. No boiling is observed in the simulations performed here.

Absolute fission yields in the fast neutron induced fission of 233U by track etch combined with gamma-ray spectrometry

The absolute fission yields of twenty seven fission products were determined in the fast neutron induced fission of 233U, employing track etch in combination with gamma-ray spectrometry. The total number of fissions was measured by registering the fission tracks on a small strip of lexan, a solid state track detector. The fission products were analysed by gamma-ray spectrometry. The measured yield values were compared to the ENDF/B-VI compilation and show a good agreement.

Kinetics of Aperiodic Cascade Boosters: Speed of Operation and Safety

The results of Monte-Carlo calculations of simple spherically symmetric systems, simulating an aperiodic booster, are presented. The purpose of the calculations is to give a clear demonsration of the specific features of the kinetics of cascade reactor systems and the possibilities which the application of the cascade principle in an aperiodic booster adds to increasing safety and improving the time parameters of the pulses. Another goal is to determine the materials for the moderator which are best from the standpoint of optimizing the parameters of the booster. It is shown that for the same total number of fissions the neutron pulses in a cascade booster are generated under conditions of much greater subcriticality and are much shorter in duration than for a single-section booster. The best pulse parameters are obtained in cascade boosters with tungsten, lead, and molybdenum neutron moderators.

Effects of very low doses of ionizing radiation on the clonal life-span in Paramecium tetraurelia.

Effects of very low doses of ionizing radiation were investigated on life-span in Paramecium tetraurelia. After autogamy single cell cultures were place (1) in a control chamber with radiopermeable walls; (2) in a shielded device with 10 cm thick lead walls; (3) in a similar shielding device including a 60Co source giving a dose of 760 mrad per year at the culture level. Life-spans, expressed in days and in mean total number of fissions measured from autogamy to the death of all sublines, were about 200 fissions and 55 days in controls. Life-spans of shielded sublines were increased and lower life-spans were observed in both shielded and irradiated sublines. These findings provide new evidence for a biological effect of very low doses of ionizing radiations and show that fission potential of Paramecium tetraurelia can be affected by variations in the background radiation level.

Absolute yields of 99Mo and 111Ag in the reactor neutron induced fission of 238U

The absolute fission yields of 99Mo and 111Ag have been measured in the reactor neutron induced fission of depleted uranium employing a fission track-cum-radiochemical technique. The absolute disintegrations were determined from beta decay and Ge(Li) gamma spectrometry. The total number of fissions were measured by fission track registration on lexan solid state track detector immersed in fissile material solution. The yields for99Mo and 111Ag were found to be (6.27 ± 0.30)% and (0.046 ± 0.002)% respectively. The results are compared with the published data and the merits of the present method have been discussed.

Verification of energy release in fission from power plant calorimetrics and isotopic composition of spent fuel

The reactor of a nuclear power station was treated as a flow calorimeter to evaluate the energy released in fission from a given loading of UO 2. The changes in the inventories of U and Pu isotopes were obtained from reprocessing of the spent fuel and the corresponding total number of fissions was determined. The resulting energy released per average fission in a fuel irradiated to 11·5GWD/MTU was found to be 203·2 ± 6·0 MeV from batch isotopic method, 204·6 ± 6·3 MeV from batch composite method and 204·8 ± 7·2 MeV from 148Nd-analyses. For burnups both higher and lower than this value, the results indicate the presence of a burnup dependence somewhat greater than that expected on the basis of Pu buildup alone. The accuracy of the plant data was found to be limited by the ability to measure the steam generator feed water flow rates.

Relative measurements of the mean number of neutrons emitted per fission induced in 233u 235u, and 239pu by thermal neutrons and “fission” neutrons

Ratios have been measured of the values of v for fission induced in 233U, 235U, and 239Pu by thermal and “fission” neutrons. The method was to count the number of coincidences between fission fragments and fission neutrons simultaneously with the total number of fissions. For all three nuclides, v increases by approximately 10 pet cent in passing from the lower energy spectrum to the higher.