Spontaneous Fission Rate Research Articles

Passive fast neutron multiplicity counting system based on a symmetric spherical design

ABSTRACT The measurement results of current Fast Neutron Multiplicity Counting (FNMC) systems are susceptible to the size and shape factors of special nuclear materials under test, primarily due to the sample position or size dependence of the detection efficiencies of these systems. To reduce this effect, a symmetric spherical design for a FNMC system is proposed. Two systems made of 32 and 16 liquid scintillation detectors were simulated by Geant4, and the spatial uniformities of the detection efficiencies of the two systems were characterized. Compared with the system made of 16 liquid scintillation detectors, the detection efficiency of the system made of 32 liquid scintillation detectors is more uniform, due to the solid angle covered by 32 detectors is larger than that covered by 16 detectors. The relative deviations (RDs) of its detection efficiencies within a spherical radius of 6 cm in the system are less than ± 3.3%. Thus, a FNMC system made of 32 liquid scintillation detectors was built. Several 252Cf neutron sources were placed on the sample platform with different spatial distributions to characterize the spatial uniformity of the detection efficiencies of the system. And their spontaneous fission (SF) rates were calculated using the FNMC analytical equation. The results show that the RDs of the detection efficiencies within a radius of 6 cm in the system are less than ± 2%, and the RDs of the estimated SF rates of the neutron sources are less than ± 1.8%.

A characterization method for reprocessed spent nuclear fuel through neutron and gamma-ray multiplicity counting

It is difficult to measure neutrons and gamma rays coming directly from uranium and plutonium atoms in pyroprocessed ingot, which is of great interest to nuclear safeguards. Most neutrons of the ingot are emitted from 244Cm isotope and gamma rays from its fission products. Therefore, it is necessary to identify spent fuel characteristics such as initial enrichment, burnup and cooling time from measurement and estimate plutonium mass from calibration curve or burnup calculations.In this study, we explored the possibility of characterizing pyroprocessed ingots using neutron and gamma-ray multiplicity counting. We particularly focused on finding the possible correlation between neutron and gamma-ray multiplicity properties (singles, doubles, and triples) and spent fuel properties for U/TRU/RE preprocessed ingot. We examined neutron and gamma-ray correlation for spent fuel, using the Cascading Gamma-ray Multiplicity with Fission (CGMF) and the Fission Reaction Event Yield Algorithm (FREYA) library in MCNP6.2. Due to high spontaneous fission rate, neutron and gamma-ray multiplicity parameters did not strongly correlate with multiplication factor from the point model but was proportional to 244Cm mass. The results based on the CGMF and FREYA fission library showed the similar trends for neutron and gamma-ray cases. The sum of two multiplicity data yielded best linearity with the 244Cm effective mass. Although total neutron counting (S) alone was sufficient for effectively characterizing spent fuel due to its strong linearity, this feasibility study confirmed that, in principle, it will be more effective to calculate the sum of neutron and gamma-ray multiplicity properties to directly estimate the mass of 244Cm, rather than relying on the conventional method of calculating the multiplication factor.

Assessment of the Photon and Neutron Source Term for the NPP Krško Spent Fuel

Accurate knowledge of the fuel nuclide inventory is important after reactor shut down, during the fuel storage and subsequent reprocessing or disposal to provide adequate shielding from the photon and neutron radiation. In this paper possibility to calculate the NPP Krško photon and neutron source term with the Serpent code has been analysed. Some deficiencies in the supplied ENDF/B-VII.0 decay library have been observed. In addition, Serpent reports only spontaneous fission rates without (α, n) and (β, n) contributions. To get neutron emission, spontaneous fission rates had to be multiplied with the average number of neutrons born for each particular nuclide manually. Comparison with the Origen code has shown acceptable agreement of the ENDF/B-VII.1 results. Influence of several factors such as fuel burnup, enrichment, temperature, moderator temperature (density), soluble boron concentration, average power, and burnable absorbers has been analysed. In addition, it was demonstrated that, except for the burnup and enrichment, averaging of all other parameters is acceptable approach. IFBA fuel should be accounted for explicitly due to relative high impact on the photon and neutron emissions.

Influence of Spontaneous Fission Rates on the r-process Nucleosynthesis

The effects of spontaneous fission on r-process nucleosynthesis are investigated in the hot wind r-process scenario. We perform network calculations using three sets of spontaneous fission rates to study how the abundance pattern is shaped when different sets of fissioning nuclei are encountered by the r-process nuclear flow. The relative contributions from spontaneous fission, neutron-induced fission, and β-delayed fission to the nucleosynthesis process are studied by calculating the corresponding fission flow. We show that the relative contributions of various fission channels in r-process nucleosynthesis depend on the astrophysical conditions and fission models used. By using the spontaneous fission rates from a modified Swiatecki’s formula with isospin and blocking effects, the spontaneous fission and neutron-induced fission play an equally important role in r-process nucleosynthesis under an extreme neutron-rich astrophysical scenario with Y e = 0.1. The fissioning nuclei are located in different regions of the nuclear chart when different spontaneous fission models are used. The fission fragment distributions of fissioning nuclei in different regions have apparent diversity, which affects the mass regions where fission products are deposited, leading to the difference of the final abundance around the second r-process peak and rare-earth subpeak.

Impact of Approximations in Operating History Data on Spent Fuel Properties With Serpent 2

Abstract In this work, the effect of averaging operating history parameters such as power history, boron concentration and coolant density, and temperature on spent nuclear fuel properties was investigated. The examined properties were assembly activity, decay heat, photon emission rate, spontaneous fission rate, and the concentration of some mobile nuclides and fissile nuclides. Calculations were performed on two similar VVER-440 fuel assemblies irradiated in different positions of the core using Serpent 2. Averaging power history over the entire irradiation history had a significant effect on assembly activity, decay heat, and photon emission rate overestimating these properties approximately 70% right after irradiation. However, the effect quickly died out and after 10 years of cooling, the effect was less than 1%. If the last cycle (third cycle) was modeled accurately and the power density of only the first two cycles was averaged, the differences remained always below 1%. The effect of operating history approximations on spontaneous fission rate and the nuclide concentrations was much smaller remaining mostly below 1.5%. The sensitivity of nuclide concentrations to approximations in individual operating history parameters was dependent on the nuclide in question and no trend applying to all studied nuclides could be observed.

The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

Abstract The estimation of spent nuclear fuel source term (decay heat, reactivity, nuclide inventory, etc.) has several sources of uncertainty such as uncertainties in nuclear data, uncertainties in the operation history, and choice of calculation parameters. In this work the effect of calculation parameters is studied by estimating the source term with the built-in burnup capability of serpent. The effect of the following parameters is considered: depletion zone division, burnup steps, unresolved resonance probability table sampling, Doppler-Broadening Rejection Correction (DBRC) and energy-dependent branching ratios. As a test case, a 2D Boiling Water Reactor (BWR) fuel assembly was modeled by first running a burnup calculation followed by a decay calculation. The following source term components were considered when investigating the effect of the studied parameters: total decay heat, photon emission rate and spontaneous fission rate. In general, the differences resulting from the use of different parameter variations were small for all three studied source term components. For the decay heat largest absolute relative difference was approximately 0.6% and for the photon emission rate approximately 1.1%. For the spontaneous fission rate maximum absolute relative difference of nearly 8% was observed. For all three components the variation of the depletion zone division resulted in the largest relative differences. Clear differences were also observed for burnup step length and DBRC variations. The use of unresolved resonance probability table sampling and energy-dependent branching ratios had an insignificant effect on the studied source term components.

The Impact of Nuclear Physics Uncertainties on Interpreting Kilonova Light Curves

Recently, analysis of the optical counterpart AT2017gfo to the gravitational wave-detected neutron star merger GW170817 has suggested a promising resolution of a long-standing debate on neutron star mergers as a source of some of the heaviest elements. However, making quantitative progress in these areas requires an accounting of the uncertainties in different aspects of physics, which is input into simulations of merging compact objects and their associated phenomena, remarkably rapid neutron capture nucleosynthesis. We investigate the uncertainties from the nuclear inputs to rapid neutron capture nucleosynthesis calculations combining different theoretical nuclear mass models, spontaneous fission rates, and fission daughter product distributions on top of the experimental nuclear data. We report that such nuclear physics uncertainties typically generate at least one order of magnitude uncertainty in the nuclear heating, which leads to uncertainties in the bolometric luminosity and the inferred mass of r-process material from the kilonova light curve.

Review and Evaluation of the Spontaneous Fission Half-lives of 238Pu, 240Pu, and 242Pu and the Corresponding Specific Fission Rates

A widely applied technical measure of nuclear safeguards and nonproliferation for the nondestructive verification and mass-assay of items containing separated plutonium is passive neutron multiplicity counting. The primary source of time correlated passive neutrons is the spontaneous fission of Pu, where the 240Pu is usually the dominant isotope with 238Pu and 242Pu also contributing to a degree depending on the grade or burnup of the material. Consequently, knowledge of the 240Pu spontaneous fission rate, expressed as fission per gram per second, and the associated uncertainty is key in interpreting measurement data absolutely. The 240Pu spontaneous fission rate derives (and vice versa) from the 240Pu spontaneous fission half-life. In this work we review and reevaluate the available experimental data on the spontaneous fission (SF) half-life of the Pu isotopes 238Pu, 240Pu and 242Pu. The SF half-lives are used to compute the corresponding specific SF rates which are the traditional nuclear data parameters needed for analytical applications, especially the nondestructive assay of plutonium for safety, security and safeguards.From 7 measurements we recommend a SF half-life of (4.745 ± 0.083) × 1010y for 238Pu corresponding to a specific SF rate of (1171 ± 20) fis ⋅ s-1 ⋅ g-1.There are 16 absolute experimental determinations of 240Pu half-life. Based on this evaluation, we find that the weighted mean of the 16 half-life determinations is (1.1608 ± 0.0091) × 1011y corresponding to a specific SF rate of (474.7 ± 3.7) fis ⋅ s−1 ⋅ g−1, where the relative uncertainty of about 0.78% is the external standard error.From 8 measurements we recommend a SF half-life of (6.766 ± 0.037) × 1010y for 242Pu corresponding to a specific SF rate of (807.7 ± 4.4) fis ⋅ s−1 ⋅ g−1. Finally, we conclude that there is a need for more experiments on all three plutonium nuclides using new techniques that have emerged in recent years.

Modeling Kilonova Light Curves: Dependence on Nuclear Inputs

Abstract The mergers of binary neutron stars, as well as black hole–neutron star systems, are expected to produce an electromagnetic counterpart that can be analyzed to infer the element synthesis that occurred in these events. We investigate one source of uncertainties pertinent to lanthanide-rich outflows: the nuclear inputs to rapid neutron capture nucleosynthesis calculations. We begin by examining 32 different combinations of nuclear inputs: eight mass models, two types of spontaneous fission rates, and two types of fission daughter product distributions. We find that such nuclear physics uncertainties typically generate at least one order of magnitude uncertainty in key quantities such as the nuclear heating (one and a half orders of magnitude at 1 day post-merger), the bolometric luminosity (one order of magnitude at 5 days post-merger), and the inferred mass of material from the bolometric luminosity (factor of 8 when considering the 8–10 day region). Since particular nuclear processes are critical for determining the electromagnetic signal, we provide tables of key nuclei undergoing β-decay, α-decay, and spontaneous fission important for heating at different times, identifying decays that are common among the many nuclear input combinations.

Instanton-motivated study of spontaneous fission of odd- A nuclei

Using the idea of the instanton approach to quantum tunneling we try to obtain a method of calculating spontaneous fission rates for nuclei with the odd number of neutrons or protons. This problem has its origin in the failure of the adiabatic cranking approximation which serves as the basis in calculations of fission probabilities. Selfconsistent instanton equations, with and without pairing, are reviewed and then simplified to non-selfconsistent versions with phenomenological single-particle potential and seniority pairing interaction. Solutions of instanton-like equations without pairing and actions they produce are studied for the Woods-Saxon potential along realistic fission trajectories. Actions for unpaired particles are combined with cranking actions for even-even cores and fission hindrance for odd-A nuclei is studied in such a hybrid model. With the assumed equal mass parameters for neighbouring odd-A and even-even nuclei, the model shows that freezing the K {\pi} configuration leads to a large overestimate of the fission hindrance factors. Actions with adiabatic configurations mostly show not enough hindrance; instanton-like actions for blocked nucleons correct this, but not sufficiently.

Fast-neutron-induced fission cross section of Pu242 measured at the neutron time-of-flight facility nELBE

The fast-neutron-induced fission cross section of $^{242}\mathrm{Pu}$ was measured at the neutron time-of-flight facility $n$ELBE. A parallel-plate fission ionization chamber with novel, homogeneous, large-area $^{242}\mathrm{Pu}$ deposits on Si-wafer backings was used to determine this quantity relative to the IAEA neutron cross-section standard $^{235}\mathrm{U}(n,f)$ in the energy range of 0.5 to 10 MeV. The number of target nuclei was determined from the measured spontaneous fission rate of $^{242}\mathrm{Pu}$. This helps to reduce the influence of the fission fragment detection efficiency on the cross section. Neutron transport simulations performed with geant4, mcnp6, and fluka2011 are used to correct the cross-section data for neutron scattering. In the reported energy range the systematic uncertainty is below 2.7% and on average the statistical uncertainty is 4.9%. The determined results show an agreement within 0.67(16)% to recently published data and a good accordance to current evaluated data sets.

Determination of the fast-neutron-induced fission cross-section of 242Pu at nELBE

The fast-neutron-induced fission cross section of 242Pu was determined in the energy range of 0.5 MeV to 10MeV at the neutron time-of-flight facility nELBE. Using a parallel-plate fission ionization chamber this quantity was measured relative to 235U(n,f). The number of target nuclei was thereby calculated by means of measuring the spontaneous fission rate of 242Pu. An MCNP 6 neutron transport simulation was used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.

Fuel cycle analysis of Advanced Burner Reactor with breed-and-burn thorium blanket

The Seed-and-Blanket (S&B) Sodium-cooled Fast Reactor (SFR) core concept was proposed for generating a significant fraction of the core power from thorium fueled breed-and-burn (B&B) blankets without exceeding the presently verified radiation damage constraint of 200 Displacement per Atom (DPA). To make beneficial use of the excess neutrons from fast reactors, the S&B core is designed to have an elongated TRU transmuting (or “TRU burner”) seed from which over 20% of the fission neutrons leak into a subcritical thorium blanket that radially surrounds the seed. The seed fuel is recycled while the blanket operates in a once-through breed-and-burn (B&B) mode. The objective of this paper is to compare the fuel cycle performance of the S&B reactor against an Advanced Burner Reactor (ABR) and a conventional Pressurized Water Reactor (PWR). For the fast reactors (SFR: ABR and S&B) the fuel cycle performance is evaluated based on a 2-stage PWR-SFR energy system while the reference nuclear system is made of once-through PWRs.It was found that relative to the ABR, the S&B core has a lower fuel cycle cost, higher capacity factor, and comparable short-term radioactivity. The discharged seed fuel from the S&B core features lower fissile Pu-to-Pu ratio, higher 238Pu-to-Pu ratio, higher specific plutonium decay heat, higher spontaneous fission rate, and lower overall material attractiveness for weapon use. Due to the significant amount of 233U discharged from the breed-and-burn thorium fueled blankets, the S&B core has much higher long-term radioactivity and radiotoxicity. Since the thorium fueled blanket operates in the breed-and-burn mode and requires no fuel reprocessing, the discharged blanket fuel is unattractive for weapons application.Compared with a PWR, the S&B core has a lower fuel cycle cost, much lower short-term radioactivity and radiotoxicity but higher long-term values, and higher proliferation resistance for the discharged plutonium. The natural uranium utilization of the 2-stage PWR-S&B system is approximately 60% higher than that of present PWRs; it is few percent higher than that of the 2-stage PWR-ABR system. Approximately 7% of the thorium fed to the blanket is converted into energy, which makes the thorium fuel utilization approximately 12 times the utilization of natural uranium in PWRs.A comprehensive fuel cycle evaluation performed with the methodology developed by the recent U.S Department of Energy’s Nuclear Fuel Cycle Evaluation and Screening campaign concludes that the PWR-S&B system has similar fuel cycle performance characteristics as the PWR-ABR system. The S&B concept may potentially feature improved economics and resource utilization relative to the ABR.

Fast-neutron-induced fission of242Pu atnELBE

The fast neutron-induced fission cross section of 242 Pu was determined in the range of 0.5 MeV to 10 MeV relative to 235 U(n,f) at the neutron time-of-flight facility n ELBE. The number of target nuclei was calculated by means of measuring the spontaneous fission rate of 242 Pu. Neutron transport simulations with Geant4 and MCNP6 are used to correct the relative cross section for neutron scattering. The determined results are in good agreement with current experimental and evaluated data sets.

Thermal fission rates with temperature dependent fission barriers

The fission processes of thermal excited nuclei are conventionally studied by statistical models which rely on inputs of phenomenological level densities and potential barriers. Therefore the microscopic descriptions of spontaneous fission and induced fission are very desirable for a unified understanding of various fission processes. We propose to study the fission rates, at both low and high temperatures, with microscopically calculated temperature-dependent fission barriers and collective mass parameters. The fission barriers are calculated by the finite-temperature Skyrme-Hartree-Fock+BCS method. The mass parameters are calculated by the temperature-dependent cranking approximation. The thermal fission rates can be obtained by the imaginary free energy approach at all temperatures, in which fission barriers are naturally temperature dependent. The fission at low temperatures can be described mainly as a barrier-tunneling process. While the fission at high temperatures has to incorporate the reflection above barriers. Our results of spontaneous fission rates reasonably agree with other studies and experiments. The temperature dependencies of fission barrier heights and curvatures have been discussed. The temperature dependent behaviors of mass parameters have also been discussed. The thermal fission rates from low to high temperatures with a smooth connection have been given by different approaches. \item[Conclusions] Since the temperature dependencies of fission barrier heights and curvatures, and the mass parameters can vary rapidly for different nuclei, the microscopic descriptions of thermal fission rates are very valuable. Our studies without free parameters provide a consistent picture to study various fissions such as that in fast-neutron reactors, astrophysical environments and fusion reactions for superheavy nuclei.

Microscopic description of fission in neutron-rich radium isotopes with the Gogny energy density functional

Mean field calculations, based on the D1S, D1N and D1M parametrizations of the Gogny energy density functional, have been carried out to obtain the potential energy surfaces relevant to fission in several Ra isotopes with the neutron number 144 $\le$ N $\le$ 176. Inner and outer barrier heights as well as first and second isomer excitation energies are given. The existence of a well developed third minimum along the fission paths of Ra nuclei, is analyzed in terms of the energetics of the "fragments" defining such elongated configuration. The masses and charges of the fission fragments are studied as functions of the neutron number in the parent Ra isotope. The comparison between fission and $\alpha$-decay half-lives, reveals that the former becomes faster for increasing neutron numbers. Though there exists a strong variance of the results with respect to the parameters used in the computation of the spontaneous fission rate, a change in tendency is observed at N=164 with a steady increase that makes heavier neutron-rich Ra isotopes stable against fission, diminishing the importance of fission recycling in the r-process.

Evaluation of Natural Radioactivity Levels for Structural Material Used in the Construction of the Neutrino Detector

In this work, the evaluation of natural radioactivity and spontaneous fission rates was performed for 8 nuclides from the natural radioactive 238U, 235U and 232Th decay chains. For this purpose, three samples of structural materials of the neutrino detector, i.e. aluminum, titanium and glass were analyzed by gamma spectroscopy and by neutron activation analysis to quantify a specific radioactivity of the samples. According to the results of this investigation, glass and aluminum samples have maximum values of the mean uranium concentrations 7.3(7) × 10-4% and 3.1(6) × 10-5%, respectively, while the lowest value for mean concentration of the uranium was found in titanium samples to be 4.7(3) × 10-6%. Aluminum sample had maximum values of the mean thorium concentrations, 2.5(8) × 10-3%, while the lowest value for mean concentration of the thorium was found in titanium samples to be 6.2(3) × 10-7%.

The optimum choice of gate width for neutron coincidence counting

In the measurement field of international nuclear safeguards, passive neutron coincidence counting is used to quantify the spontaneous fission rate of certain special nuclear materials. The shift register autocorrelation analysis method is the most commonly used approach. However, the Feynman-Y technique, which is more commonly applied in reactor noise analysis, provides an alternative means to extract the correlation information from a pulse train. In this work we consider how to select the optimum gate width for each of these two time-correlation analysis techniques. The optimum is considered to be that which gives the lowest fractional precision on the net doublets rate. Our theoretical approach is approximate but is instructional in terms of revealing the key functional dependence. We show that in both cases the same performance figure of merit applies so that common design criteria apply to the neutron detector head. Our prediction is that near optimal results, suitable for most practical applications, can be obtained from both techniques using a common gate width setting. The estimated precision is also comparable in the two cases. The theoretical expressions are tested experimentally using 252Cf spontaneous fission sources measured in two thermal well counters representative of the type in common use by international inspectorates. Fast accidental sampling was the favored method of acquiring the Feynman-Y data. Our experimental study confirmed the basic functional dependences predicted although experimental results when available are preferred. With an appropriate gate setting Feynman-Y analysis provides an alternative to shift register analysis for safeguards applications which is opening up new avenues of data collection and data reduction to explore.

Assessment and reduction of proliferation risk of reactor-grade plutonium regarding construction of ‘fizzle bombs’ by terrorists

The approximately 23.7wt% 240Pu in reactor-grade plutonium denatures the 239Pu to the extent that it cannot fuel high yield nuclear weapons. 240Pu has a high spontaneous fission rate, which increases the spontaneous neutron flux within the fuel. When such a nuclear weapon is triggered, these neutrons cause the nuclear fission chain reaction to pre-detonate which blows the imploding fuel shell apart before the designed level of compression and reactivity could be attained, thereby greatly reducing the average energy yield of such “fizzle” bombs. Therefore reactor-grade plutonium is normally viewed as highly proliferation resistant. In this article the literature on the proliferation resistance of reactor-grade plutonium and on the mechanism and effect of fizzle bombs is reviewed in order to test this view. It is shown that even very low yield fizzle bombs, exploded in urban areas, would still cause serious blast damage as well as radioactive contamination. Combined with the high levels of induced terror, fizzle bombs might thus be attractive psychological weapons for terrorists. Therefore reactor-grade plutonium may not be sufficiently proliferation resistant against nuclear terrorism. However, denaturisation with more than 9% 238Pu produces high levels of decay heat which will melt or explode the high explosives around uncooled implosion type weapons, rendering them useless. Unfortunately, reactor-grade Pu contains only 2.7% 238Pu and is thus not sufficiently proliferation resistant in this respect. It is also shown that the associated neptunium poses a substantial proliferation risk.In the present study strong improvement of the proliferation resistance was demonstrated by simulation of incineration of reactor-grade plutonium in the 400MWth Pebble Bed Modular Reactor Demonstration Power Plant. Results for modified fuel cycles, aimed at transmutating 237Np to 238Pu are also reported. However, these modifications increased the disloaded heavy metal mass, thereby substantially increasing the radiotoxicity of the spent fuel. Therefore this intervention is not recommended. 237NP should thus rather be incinerated it in fast reactors, light-water reactors or CANDU reactors.

Effect of spontaneous fission models on the production of cosmochronometer nuclei in the r-process

The dependence of the yields of cosmochronometer nuclei on the spontaneous fission model has been considered. It has been shown that the spontaneous fission rates estimated within a phenomenological model constructed on the basis of the calculations of the spontaneous fission rates for superheavy nuclei within the macroscopic-microscopic model are in good agreement with the observations.