Neutron Transmission Probability Research Articles

Improved Collision Probability Analysis of the Double-Heterogeneity Problem Based on Matrix Chord Length Correction

Dispersion fuel exhibits excellent safety performance and effectively reduces the risk of radioactive leakage, making it widely applied in high-temperature gas-cooled reactors (HTGRs) and other advanced nuclear reactors. The presence of stochastic media in dispersion fuel leads to the challenging double-heterogeneity problem in neutron transport calculations. Hébert proposed a collision probability analysis model for treating stochastic media, which has been implemented in the DRAGON5 code. As one important basis of derivation, it is assumed in the Hébert model that the neutron transmission probability is identical to the neutron escaping probability in matrix material. In this paper, it is figured out that the assumption is not rigorous for realistic stochastic media. Then, an improved approach based on the Hébert model is proposed to take into account the realistic chord length distribution as well as to ensure the conservation and reciprocity of collision probabilities. The proposed methodology has been implemented in the HTGR lattice physics code XPZ. By numerical analysis against Monte Carlo reference solutions, it is demonstrated that the improved Hébert model with chord length correction gives good accuracy for addressing realistic double-heterogeneity problems.

The neutron streaming correction method in the homogenization of Pebble-bed HTGR reflector

In pebble-bed high temperature gas-cooled reactors (HTGRs), the outside graphite reflector generally contains void regions, including the boring channels for control rods, absorber balls and coolant flow. Deterministic reactor physics calculation of HTGR is usually based on regular geometries, which requires pre-homogenization to the materials with void regions. However, the volume-weighted homogenization to the HTGR reflector can greatly overestimate the keff, due to the neutron streaming effect in the reflector channels. This paper proposes a convenient method of correcting the neutron streaming effect in the HTGR reflector, by adjusting the homogenized transport cross sections based on preservation of the neutron transmission probabilities. The proposed method has been implemented in the PANGU code for pebble-bed HTGR physics analysis and fuel cycle simulations. Numerical results suggest that the proposed method can effectively reduce the discrepancy in the homogenization of the HTGR reflector.

New concept of the buildup factor in bent ducts

Abstract A major problem confronting the radiation shielding designer is the accurate determination of neutron streaming through various penetrations in walls, ducts and mazes. The previous studies on neutron transmission were performed through empty ducts. The aim of this work is to evaluate the neutron transmission probability through a filled bent duct and the proposition of a new concept of the buildup factor for neutrons in multilegged ducts. An angular biaising technique is used in the Monte Carlo simulations to accelerate the calculation convergence. Results are first compared with those obtained by the MCNPX code. For an empty bent duct, the neutron transmission is only due to the neutron reflection on the duct wall. For a filled duct, the major contribution is due to the scattering on the atoms filling the duct.

Neutron transmission probability through a revolving slit for a continuous source and a divergent neutron beam

An analytical calculation for the transmission probability of neutrons travelling through a revolving slit is presented here. For the first time, two effects have been taken into account in the same approach, that is, on the one hand, the fact that the neutron beam might be divergent and, on the other hand, that neutrons from a continuous source can arrive at different times at the chopper. Furthermore, the neutron distribution at a given distance behind the chopper has been calculated and these theoretical results have been compared with simulated data obtained with the VITESS simulation program. The theoretical and the simulated curves are in good agreement.

Neutron Transmission of a Single-Crystal MgO Filter

The neutron transmission probability through a single-crystal MgO filter has been measured at both liquid-nitrogen (77 K) and room (300 K) temperatures, as a function of wavelength, using a pulsed neutron source. The data show that a cooled MgO filter is superior to room-temperature sapphire for both thermal and cold neutrons, principally because the absorption probability is reduced by a factor of three at long wavelengths.

The Albedo effect on neutron transmission probability

The aim of this study is to evaluate the albedo effect on the neutron transmission probability through slab shields. For this reason we have considered an infinite homogeneous slab having a fixed thickness equal to 20 λ (λ is the mean free path of the neutron in the slab). This slab is characterized by the factor P s (scattering probability) and contains a vacuum channel which is formed by two horizontal parts and an inclined one (David, M. C. (1962) Duc and voids in shields. In Reactor Handbook, Vol. III, Part B, p. 166). The thickness of the vacuum channel is taken equal to 2λ. An infinite plane source of neutrons is placed on the first face of the slab (left face) and detectors, having windows equal to 2λ, are placed on the second face of the slab (right face). Neutron histories are sampled by the Monte Carlo method (Booth, T. E. and Hendricks, J. S. (1994) Nuclear Technology 5) using exponential biasing in order to increase the Monte Carlo calculation efficiency (Levitt, L. B. (1968) Nuclear Science and Engineering 31, 500–504; Jehouani, A., Ghassoun, J. and Abouker, A. (1994) In Proceedings of the 6th International Symposium on Radiation Physics, Rabat, Morocco) and we have applied the statistical weight method which supposed that the neutron is born at the source with a unit statistical weight and after each collision this weight is corrected. For different values of the scattering probability and for different slopes of the inclined part of the channel we have calculated the neutron transmission probability for different positions of the detectors versus the albedo at the vacuum channel-medium interface. Some analytical representations are also presented for these transmission probabilities.

The neutron channeling phenomenon

Shields, used for protection against radiation, are often pierced with vacuum channels for passing cables and other instruments for measurements. The neutron transmission through these shields is an unavoidable phenomenon. In this work we study and discuss the effect of channels on neutron transmission through shields. We consider an infinite homogeneous slab, with a fixed thickness (20λ, with λ the mean free path of the neutron in the slab), which contains a vacuum channel. This slab is irradiated with an infinite source of neutrons on the left side and on the other side (right side) many detectors with windows equal to 2λ are placed in order to evaluate the neutron transmission probabilities (Khanouchi, A., Aboubekr, A., Ghassoun, J. and Jehouani, A. (1994) Rencontre Nationale des Jeunes Chercheurs en Physique. Casa Blanca Maroc; Khanouchi, A., Sabir, A., Ghassoun, J. and Jehouani, A. (1995) Premier Congré International des Intéractions Rayonnements Matière. Eljadida Maroc). The neutron history within the slab is simulated by the Monte Carlo method (Booth, T. E. and Hendricks, J. S. (1994) Nuclear Technology 5) and using the exponential biasing technique in order to improve the Monte Carlo calculation (Levitt, L. B. (1968) Nuclear Science and Engineering 31, 500–504; Jehouani, A., Ghassoun, J. and Aboubker, A. (1994) In Proceedings of the 6th International Symposium on Radiation Physics, Rabat, Morocco). Then different geometries of the vacuum channel have been studied. For each geometry we have determined the detector response and calculated the neutron transmission probability for different detector positions. This neutron transmission probability presents a peak for the detectors placed in front of the vacuum channel. This study allowed us to clearly identify the neutron channeling phenomenon. One application of our study is to detect vacuum defects in materials.

The vacuum geometry effect on neutron transmission and spatial resolution of neutron transmission

Frequently, shields used against radiation contain some vacuum channels. We have therefore considered an infinite slab with a fixed thickness (thickness 20 λ with λ the mean free path of the neutron in the slab) and an infinite plane source of neutrons which arrived on the left side of the slab; transmitted neutrons through the slab to its right side are detected by finite detectors having windows equal to 2λ. This slab contains a vacuum channel. This channel has many legs with several horizontal parts. We used the Monte Carlo method for sampling the neutron history in the slab with a spatial biasing technique in order to accelerate the calculation convergence (Levitt, L. B. (1968) Nuclear Science and Engineering 31, 500–504; Jehouani, A., Ghassoun, J. and Aboubker, A. (1994) In Proceedings of 6th International Symposium on Radiation Physics, Rabat, Morocco). We studied the effects of the angle position and the number of horizontal parts of the channel on the neutron transmission. We have studied the effect of the vacuum channel opening (Artigas, R. and Hungerford, H. E. (1969) Nuclear Science and Engineering 36, 295–303) on the neutron transmission; for several values of this opening we have calculated the neutron transmission probability for each detector position. This study allowed us to determine the optimal conditions of vacuum geometries to improve protection against neutrons. In the second part we considered a shield which consists of a slab and a two-legged vacuum channel with two horizontal parts. The spatial distribution of neutrons transmitted through the protection screen was determined. This distribution shows two peaks. The study was made for different distances between the two horizontal parts. We have determined the smallest distance between the two horizontal parts for which the two peaks can be resolved.

Improvements in Calculating Neutron Transmission Probabilities in Unit-Cell Interface-Current Codes

A standard assumption used in unit-cell interface-current codes is that neutrons enter each spatial region with an isotropic angular distribution. The physical interpretation of this assumption is discussed, and the magnitude of the error introduced by it is shown for a range of practical unitcell geometries. An improvement on the calculation of first-flight transmission probabilities for onedimensional unit cells based on limiting the neutron source angular distribution to physically possible neutron flight directions is then presented. For three-region problems, one additional calculation of a revised outer region transmission probability is sufficient for determining all the revised transmission probabilities of interest. Calculation of the revised transmission probabilities requires only minor coding changes and eliminates the improper angular redistribution of neutrons at the region boundaries.

Neutron Transmission Probabilities Across Control Blades Filled with Round Tubes: Formulation and Accuracy

Adequate knowledge of the transmission probabilities of neutrons through control blades filled with round tubes containing absorbing material is required in reactor design. Starting from a strict f...

制御棒のEpithermal Neutron吸収効果の計算

The average neutron transmission probability is necessary in computing the control rods epithermal effectiveness and this calculation has been done for various control rods at normal temperature from the fundamental nuclear cross sections. Those for which we have computed are; 1 Materials (a) Hf, Ag, In and Ag-In-Cd alloy (80:15:5 w/o) (b) Eu2O3 in stainless steel 2 Thickness 0.125, 0.25, 0.50 and 0.75cm for (a) group 2.5, 5.0, 10.0 and 20.0 w/o with 0.50cm thickness for (b) group By comparing resultant values with those obtained experimentally by Bach and Way in Knolls Lab., for Hf, Ag and In, satisfactory agreement has been found in case we apply a modification, caused by deviation from the 1/E spectrum in the epithermal region, to the average transmission probability. We have reached to a conclusion that the modification due to the reactor spectrum is quite essential and that with applying this we can expect our results accurate enough to use.

Neutron transmission probability through a curved revolving slit

The formulas of a neutron burst transmitted by a revolving curved slit are analytically derived and are then applied to determine the distribution of the neutrons of the burst, after they have travelled a given distance.