Investigation of level density parameter dependence for some 233U, 235U, 237U, 239U, 249Cf, 251Cf, 237Pu and 247Cm nuclei in neutron fission cross sections with the incident energy up to 20 MeV

Level density models have increasing importance to gain more in-depth into the nature of nuclear reactions. Many novel and advanced medical application use radioisotopes, which are produced with nuclear reactions. In this study, the effect of the level density parameters of the nucleus on the cross sections of neutron-fission reactions for 233U, 235U, 237U, 239U, 249Cf, 251Cf, 237Pu and 247Cm nuclei were investigated for up to 20 MeV neutrons. TALYS 1.8 software was used to calculate the cross-sections of neutron-fission reactions for different level density parameters. The calculations were compared with the EXFOR nuclear data library and the level density parameters, and the closest fit were searched. As outputs of the study, the effect of selection of level density parameter on cross section calculations was observed. The theoretically obtained data were compared with the experimental data taken from the literature. The results are presented graphically for better interpretation.

AN IMPROVED ENERGY DEPOSITION MODEL IN MPACT AND EXPLICIT HEAT GENERATION COUPLING WITH CTF

The default energy deposition model in the CASL neutronics code MPACT assumes all fission energy is deposited locally in fuel rods. Furthermore, equilibrium delayed energy release is assumed for both steady-state and transient calculations. These approximations limit the accurate representation of the heat generation distribution in space and its variations over time, which are essential for power distribution and thermal-hydraulic coupling calculations. In this paper, an improved energy deposition model is presented in both the spatial and time domains. Spatially, the energy deposition through fission, neutron capture, and slowing-down reactions are explicitly modeled to account for the heat generation from all regions of a reactor core, and a gamma smearing scheme is developed that utilizes the gamma sources from neutron fission and capture. In the time domain, the delayed energy release is modeled by solving an additional equation of delayed heat emitters, similar to the equation of delayed neutron precursors. To allow the explicit heat generation coupling, the interfaces between MPACT and CTF were updated to transfer separate heat sources for different material regions (fuel, clad, moderator and guide tube). The results show that the distributions of the energy deposition between MPACT and MCNP agree very well for various 2-D assembly and quarter-core problems without TH feedback. The MPACT/CTF coupled calculation for the hot full power quarter-core case exhibited a reduced peak pin power by 2.3% and a reduced peak fuel centerline temperature by 17 K when using the explicit energy deposition and heat transfer. The new model also shows a maximum 100 pcm keff effect on assembly depletion problems and an increased overall energy release by 7% in a PWR reactivity-initiated accident (RIA) problem.

INLINE THERMAL AND XENON FEEDBACK ITERATIONS IN MONTE CARLO REACTOR CALCULATIONS

In this work, we describe a method for converging nonlinear feedback during the convergence of the neutron fission source in a Monte Carlo reactor simulation. This approach involves updating feedback physics during discard batches in the Monte Carlo simulation rather than fully (or partially) converging the neutronics prior to the nonlinear update. This approach is demonstrated for a single PWR pin with thermal feedback and with both thermal and xenon feedback. Converging these feedbacks inline with the fission source is shown to have the benefit of reducing numerical instability by effectively underrelaxing the tallied quantities in the Monte Carlo simulation and improving computational performance by converging feedback within (or near to) the number of discard batches required to converge the fission source even without any feedback.

PIN POWER CALCULATION SCHEME FOR REACTOR PRESSURE VESSEL FAST NEUTRON FLUENCE ESTIMATION

The estimation of the neutron fluence at the Reactor Pressure Vessel (RPV) is classically carried out by a two-step approach. The first step is to estimate the full core neutron source term whether the second step of the calculation consists in the transport of neutrons from the core (source term) to the RPV using the neutron fission distribution determined in the previous step. For this purpose, the neutron fission distribution is to be accurately determined at the fuel pin level for the assemblies on the border of the core. To achieve this goal, two methods are evaluated in this study. The first method considered is a full core 2D Monte Carlo calculation using the MNCP6 code. The second method is based on a deterministic approach using the CASMO5 multi-segment option, allowing a full 2D transport calculation at the pin level with an expected accuracy similar to a stochastic method. The comparison of the two methods shows an overall good agreement with differences within the statistical uncertainty for different cores: homogeneous UOX core, mixed UOX-MOX loading and the effect of the hafnium rods used in the assemblies in the periphery of the core. The modelling limitation and the associated calculational time are discussed for the comparison of the two approaches.

Evaluation of the neutronic performance of a fast traveling wave reactor in the Th-U fuel cycle

The possibility for all of the uranium or thorium fuel to be used nearly in full is expected in traveling wave reactors. A traveling wave reactor core with a fast neutron spectrum in a thorium-uranium cycle has been numerically simulated. The reactor core is shaped as a rectangular prism with a seed region arranged at one of its ends for the neutron fission wave formation. High-enriched uranium metal is used as the seed region fuel. Calculated power density dependences and concentrations of the nuclides involved with the transformation chain along the core at a number of time points have been obtained. The results were graphically processed for the clear demonstration of the neutron fission wave occurrence and transmission in the reactor. The obtained power density dependence represents a soliton (solitary wave) featuring a distinct time repeatability. Neutron spectra and fission densities are shown at the initial time point, when no wave has yet formed, and at the time of its formation. The wave rate has been calculated based on which the reactor life was estimated. The fuel burn-up has been estimated the ultra-high value of which makes the proposed reactor concept hard to implement. The burn-up of most of both the raw material and the fissile material it produces indicates a high potential efficiency of the developed reactor concept in terms of fuel utilization and nuclear nonproliferation.

Development of a “fission-proxy” method for the measurement of 14-MeV neutron fission yields

Relative fission-yield measurements were made for 50 fission products from 25.6±0.5 MeV α-induced fission of Th-232. Quantitative comparison of these experimentally measured yields with the evaluated fission yields from 14-MeV neutrons on U-235 demonstrates the application of the Bohr-independence hypothesis for measuring fission yields. As optimum particle-target configurations may be impossible or compromised at a given facility, this new approach, fission-proxy, allows the measurement of fission yields for a given compound nucleus from an alternate reaction pathway since formation and subsequent decay are independent processes.

Impact of nuclear inertia momenta on fission observables

Fission is probably the nuclear process the less accurately described with current models because it involves dynamics of nuclear matter with strongly coupled manybody interactions. It is thus diffcult to find models that are strongly rooted in good physics, accurate enough to reproduce target observables and that can describe many of the nuclear fission observables in a consistent way. One of the most comprehensive current modeling of the fission process relies on the fission sampling and Monte-Carlo de-excitation of the fission fragments. This model is implemented for instance in the FIFRELIN code. In this model fission fragments and their state are first sampled from pre-neutron fission yields, angular momentum distribution and excitation energy repartition law then the decay of both initial fragments is simulated. This modeling provides many observables: prompt neutron and gamma fission spectra, multiplicities and also fine decompositions: number of neutrons emitted as a function of the fragment mass, spectra per fragments, etc. This model relies on nuclear structure databases and on several basic nuclear models describing for instance gamma strength functions or level densities. Additionally some free parameters are still to be determined, namely two parameters describing the excitation energy repartition law, the spin cutoff of the heavy and light fragments and a rescaling parameter for the rotational inertia momentum of the fragments with respect of the rigid-body model. In the present work we investigate the impact of this latter parameter. For this we mainly substitute the corrected rigid-body value by a quantity obtained from a microscopic description of the fission fragment. The independent-particle model recently implemented in the CONRAD code is used to provide nucleonic wave functions that are required to compute inertia momenta with an Inglis-Belyaev cranking model. The impact of this substitution is analyzed on different fission observables provided by the FIFRELIN code.

Detection of Radon-Progeny and Other Alpha-Emitting Radionuclides in Air Using Tensioned Metastable Fluid Detectors

Alpha radiation emitting radon (Rn) gas seepage into homes in the USA leads to over 21,000 annual lung cancer deaths (according to the US-Environmental Protection Agency, EPA) leading to mandatory monitoring for Rn throughout the USA. In the nuclear industry alpha emitting radionuclides in air (e.g., in spent fuel reprocessing) also constitute a major safety and security-safeguards related issues. Purdue University, along with Sagamore Adams Laboratories LLC, is developing the tensioned metastable fluid detector (TMFD) technology for general-purpose alpha-neutron-fission spectroscopy. This paper focuses on rapid, high-efficiency detection of Rn and progeny in air using the novel TMFD technology; Rn and progeny isotopes in air are sparged through the TMFD detection fluid (to entrap the radioactive gas), which is then placed under a metastable state. Through tailoring the metastable fluid state, an audible and visible cavitation detection event is created and readily detected from transient bubble formation. Changing the tensioned state allows for the spectroscopic differentiability of Rn and its daughters which can be used to actively measure the equilibrium between the parent and daughter products. Such a technique can also be used to monitor the atmosphere in critical nuclear facilities for contamination from other alpha emitting isotopes (e.g., Pu, Cm, U...). TMFDs offer the unique ability for high intrinsic efficiency (>95%) alpha-neutron-fission fragment detection, while remaining blind to background beta and gamma radiation (qualified to >3.8×108 Bq m−3 using a dissolved 32P beta source, and also via gammas from a 53 R/hr 137Cs gamma source). Immunity to beta and gamma is beneficial for the discrimination of buildup of beta-emitting Thoron and Rn progeny in the detection fluid allowing for reusability. This paper will discuss the research results pertaining to detection of Radon and progeny in air, for concentrations between 74 Bq m−3 (2 pCi/L) and 740 Bq m−3 (20 pCi/L). The system measures a radon concentration between these levels to within ±15% intrinsic relative error (IRE) within 24 hours meeting the standards outlined by the American Association of Radon Scientists and Technicians-National Radon Proficiency Program (AARST-NRPP) Device Evaluation Program. Precision evaluation was also performed and the relative standard deviation defined by the AARST-NRPP was <5% exceeded the requirement of 25%. Ambient temperature effects were assessed at 10 °C and at 27 °C, which revealed a large increase in collection efficiency with decreasing sampling temperature and slight increase with increasing sampling temperature. Temperature effects on sensitivity thresholds and volumetric expansion were measured and used to compensate for variability in temperature over time. Blind testing with the help of Bowser-Morner Radon Reference Laboratory was performed and succeeded in accurately determining the Rn in air concentration to within 20% within only 6h of sampling. Finally, a 48-hour based collection time has also been developed for use in dwellings where Rn in air concentrations may vary in a day. Overall, the reproducibility and precision of TMFD technology is found to allow for an efficient, cost-effective, reliable, and environmentally friendly means of Rn and progeny detection, and by extension for use for general actinide in air monitoring for the nuclear industry.