Neutron Streaming Research Articles

Experiments and Calculations on Neutron Streaming through Bent Ducts.

Neutron spectra in a cylindrical straight duct and in bent ducts with angles of 30°, 60°and 90° have been measured by the multiple foil activation and thermoluminescence dosimetry methods. Two-dimensional discrete ordinates and three-dimensional Monte Carlo calculations are executed, and the results are compared with the measurements. The flow rate at the duct entrance calculated by the DOT3.5 code is underestimated by approximately 30%, due to a conversion of the core and reflector geometry from XY to RZ geometry. The fast neutron flux in the ducts is underestimated by 20% by the MORSE-SGC/S code due to a too coarse angular mesh of the source, which does not properly represent the actual angular distribution of the fast flux, which is highly peaked forwardly into the ducts. The thermal neutron flux was overestimated by the Monte Carlo calculation. A method is proposed to calculate the angular distribution of the flow rate at the duct entrance and to calculate the source strength and the angular distribution of the flow rate at the entrance of the second leg of the duct. The results are compared with those of the transport calculations. Generally, the agreement is quite satisfactory.

The Neutron Streaming Term in a Generalized Coordinate System

The vectorial form of the neutron streaming term used in the Boltzmann equation has been derived. With the aid of this vectorial form, the explicit expression of the neutron streaming term in a generalized coordinate system can readily be obtained. In this paper examples of applications in the orthogonal curvilinear and the elliptical-elliptical toroidal coordinate systems are given for illustration.

A three dimensional calculation of neutron streaming through ITER tokamak pumping ducts with the Monte Carlo code TRIPOLI-2

In the frame of the ITER tokamak project, the streaming of neutrons through pumping ducts up to the properly so called pumping system is studied. The gas evacuation device of the ITER plasma consists of a set of vacuum pumps which are located in a room which is outside the main machine building. These pumps receive the exhaust gas through several pumping ducts with a cross section of about four square meters and a length of about ten meters. Although insensitive to the magnetic field, the 14 MeV neutrons from plasma D-T thermonuclear reactions can penetrate in the divertor and reach the room pumping device by propagation through the bent ducts. Differents components of this system, such as the bellows, turbomolecular pumps, etc., are irradiated and that raises radiation problems. In this study we determine, by using 3D Monte Carlo transport code TRIPOLI-2, neutron fluxes, dose rates and heatings due to neutrons which have streamed out the plasma through the bent ducts, at several points of the pumping room. Results show the neutron flux attenuation reachs a factor 10 −5 from plasma chamber to the pumping hall; the neutron heatings are estimated to 1.9 × 10 −3 W/cm 3 in bellow stainless steel at duct entrance, and 8 × 10 −7 W/cm 3 in the turbopumping stainless steel structure, inside pumping hall. The neutron fluxes obtained will be used to compute gamma source produced by radiative, inelastic process and gamma rays from formed activation products. Then, the knowledge of gamma source will allow to compute gamma dose rate and heating. The dose rates and heatings obtained will contribute to the definition of the ITER pumping system technical options and to establish pumping hall access conditions, also.

Application of forward and adjoint Monte Carlo coupling technique to detector system designs

Forward and adjoint Monte Carlo coupling technique has been developed for analyzing neutron streaming in a system with large geometry. Particles (neutron and adjoint particle) are scored by surface type estimators such as the next event surface crossing estimators and the boundary crossing estimators. The detector response is calculated by folding the calculated neutron and adjoint angular fluxes. The reliability and efficiency for this method were studied by solving a sample problem of neutron streaming through narrow sodium pipe embedded in an iron shield. This method turned out to give a figure of merit several times better than the conventional method. The applicability of the method to detector system design has been demonstrated by calculating the signal to noise ratio for the fuel failure detector with delayed neutron detection method, which is located behind the reactor shield of concrete. This method gives an advantage in clarifying the spatial channels for neutron streaming.

Evaluation of Neutron Streaming in Fast Breeder Reactor Fuel Subassembly by Double Heterogeneous Modeling

Neutron streaming in a fast breeder reactor fuel subassembly caused by the double heterogeneity of the pin structure and the wrapper tube structure is estimated by double heterogeneous modeling. Th...

Fast Neutron Streaming through Duct in Concrete Shields

Measurements on the neutron and γ-ray attenuation in multi-legged air filled ducts have been carried out using a Cockroft-Walton type neutron generator. The measured spectra were obtained with an NE-213 liquid scintillator using pulse shape discrimination to resolve neutron and γ-ray pulse height data and using a spectral unfolding code to convert these data to energy spectra. Results are given for two rectangular duct geometries. A series of integral measurements have also been carried out with a long counter for fast neutrons and a radiation survey meter to measure the γ-ray dose. The biological dose arising due to the streaming of neutrons was measured with a rem counter. The results show the relative decrement of the integral counts along the axis of the duct. The spectral measurements reveal the presence of resonant windows of oxygen present in the concrete medium. Also the 14 MeV component is found to be dominant in the first leg of the duct whereas degraded neutrons dominate in the second leg. The ...

Experimental Verification of Calculated Neutron Streaming in the Bottom Reflector and Core Support Regions of a Large-Scale High-Temperature Gas-Cooled Reactor Design Concept

The analysis of an Oak Ridge National Laboratory Tower Shielding Facility (TSF) experiment in which measurements were made of neutrons streaming through a mockup of a section of the lower core support structure of a large-scale high-temperature gas-cooled reactor (HTGR) design concept is described. The analysis was performed with the same calculational methods used for an analysis of the HTGR design itself, the purpose of the experiment being to provide data against which the validity of the calculational methods could be tested. Also summarized are the HTGR design calculation results; how they affected the design and objectives of the TSF experiment is described. Comparisons of the neutron detector responses observed in the experiment with calculated responses showed satisfactory agreement in most cases, and the implications of these results for the HTGR shield design are highlighted.

Comments on “Neutron Streaming through Gaps in Fusion Reactor Shielding”

Comments on “Neutron Streaming through Gaps in Fusion Reactor Shielding”

Experiments and Calculations of 14 MeV Neutron Streaming through Multi-Layered Slit Assembly

Experiments and Calculations of 14 MeV Neutron Streaming through Multi-Layered Slit Assembly

Development ofSn-AMC-AMC Coupling Technique and Its Application to Analysis of Neutron Streaming Data of “JOYO”

Development of<i>Sn</i>-AMC-AMC Coupling Technique and Its Application to Analysis of Neutron Streaming Data of “JOYO”

Development of Sn-AMC-AMC coupling technique and its application to analysis of neutron streaming data of "JOYO"

An Sn-albedo Monte Carlo-albedo Monte Carlo (Sn-AMC-AMC) coupling technique has been developed and implemented in the AMC code system MORSE-ALB in order to make an effective and accurate shielding calculation for large and complex geometry. This paper describes the feature of the present coupling method which includes new source geometry in the Sn-AMC coupling calculation. Applicability of the present method was studied by analyzing the experiments of neutron streaming in “JOYO”. The Sn-AMC-AMC coupling calculation reproduced the measured data within the accuracy of one order of magnitude. Considering that neutron fluxes attenuate by 8∼12 orders of magnitude from the core to the measuring positions and the streaming paths are complicated, the agreement can be regarded good.

Neutron Streaming Through Gaps in Fusion Reactor Shielding

Calculations have been performed using the Monte Carlo code, MORSE-CG, to determine the neutron streaming through various straight and stepped gaps between radiation shield sectors in the conceptua...

New Efficient Albedo Monte Carlo Method for Neutron Streaming Analyses

An efficient albedo Monte Carlo method newly developed has been studied by analyzing two types of experiments on neutron streaming. The method is characterized by employing the energy-angle dependent doubly differential albedos for slab, which can be calculated in a short computer time with a one-dimensional transport theory, such as the Sn method and more efficient invariant imbedding method. This paper describes the features of the present albedo Monte Carlo method, including fundamental formulas. In the analyses of the neutron streaming experiments, the calculated results agreed with the measured data within a factor of 2 for a benchmark experiment at the YAYOI reactor and within a factor of 3 for an SNR sodium duct mock-up experiment. It is concluded that the present albedo Monte Carlo method is practical and applicable to the reactor shielding analysis concerning radiation streaming.

Application of the Direct Statistical Approach on a Multisurface Splitting Problem in Monte Carlo Calculations

The direct statistical approach was used recently to obtain analytic expressions for the second moment and the average time per source particle in a general case of multisurface geometry splitting. A practical algorithm is given for the calculation of the second moment as a function of the splitting parameters on the surfaces. The algorithm is used to calculate an explicit expression for the efficiency in a problem involving streaming of neutrons in a straight section of a concentric coolant pipe of a fast reactor. The results are used to obtain the optimum splitting range on four surfaces.

Neutron Streaming through Narrow and Long Annulus

Neutron Streaming through Narrow and Long Annulus

Neutron streaming analysis for helical geometry system, heliotron-H fusion power reactor.

Nuclear analysis was carried out for the Heliotron-H reactor. Due to the helical geometry of the system, 14 MeV neutron streaming inherent in the reactor exists; some of source neutrons generated in the plasma region can see the coil shield surface in a direct line-of-sight manner and stream through blanket regions without collisions. With an appropriate coordinate transformation from the helical geometry, analysis on the neutron streaming was made possible by a method coupling the three-dimensional Monte Carlo code and two-dimensional discrete-ordinates transport code. For each radiation response parameter in the helical magnet, the streaming neutrons contribute 40–50%. This fact strongly suggests the need of taking account of the contribution from the streaming neutrons to provide an adequate shielding for the helical magnet.

Neutron streaming through narrow and long annulus.

Neutron streaming in a very narrow and long stainless steel annulus which was 3 mm wide and 2,480 mm long, was measured by TLDs and a neutron dose rate meter at the fast neutron source reactor YAYOI. The attenuation was very strong near the entrance and became very weak beyond 900 mm. The weak attenuation was due to the streaming of neutrons through the annulus. The two-dimensional discrete ordinates transport calculation with a forward biased S418 quadrature set failed to predict the attenuation beyond the point where the most-forward ray from the entrance intersected the wall of the annulus. It was practically impossible to analyze the streaming problem by the Sn code with more detailed Sn quadratures. The analytic expression describing the direct component agreed fairly well with the experiment far from the entrance of the annulus.

New efficient albedo Monte Carlo method for neutron streaming analyses.

An efficient albedo Monte Carlo method newly developed has been studied by analyzing two types of experiments on neutron streaming. The method is characterized by employing the energy-angle dependent doubly differential albedos for slab, which can be calculated in a short computer time with a one-dimensional transport theory, such as the Sn method and more efficient invariant imbedding method. This paper describes the features of the present albedo Monte Carlo method, including fundamental formulas. In the analyses of the neutron streaming experiments, the calculated results agreed with the measured data within a factor of 2 for a benchmark experiment at the YAYOI reactor and within a factor of 3 for an SNR sodium duct mock-up experiment. It is concluded that the present albedo Monte Carlo method is practical and applicable to the reactor shielding analysis concerning radiation streaming.

Comparison of numerical solutions to predict neutron streaming in shield design for reprocessing plants

Chemical plants which handle or reprocess irradiated nuclear fuel are constructed with shield walls to protect the operating staff from intense sources of radiation. These walls can contain penetrations which are designed to provide services into cells. The shield designer requires both flexible and accurate methods for the prediction of dose-rates due to neutrons streaming and scattering along these penetrations. The objective of this work is to present approximate solutions obtained from the finite element, discrete ordinate and Monte Carlo analyses of a ventilation duct. The methodology is based upon the two dimensional transport theory solutions implemented in the finite element and the discrete ordinate codes FELTRAN and DOT 4.2. The three dimensional deep penetration shielding code McBEND is used to obtain Monte Carlo solutions. The calculational model is a two-dimensional idealisation of a ventilation duct based upon the design supplied by BNFL. The first and third leg of the duct is normal to the shield surface. In addition, a bulk shielding problem containing an air-gap has been solved to obtain radiation levels due to neutron penetration as a result of a criticality incident. The results presented in this paper showed that the new numerical technique implemented in the finite element code FELTRAN, can treat air-gaps and voids in two-dimensions reasonably rigorously. Comparisons of the finite element solutions against the discrete ordinate and the Monte Carlo results showed reasonable agreement in most cases. The results obtained from FELTRAN, DOT 4.2 and McBEND are presented in graphical form. The potentials and the weaknesses of the numerical techniques implemented in these codes are discussed.

Neutron dosimetry in the three-mile island unit 2 reactor cavity with solid-state track recorders

Solid-state track recorder (SSTR) neutron dosimetry has been conducted in the Three-Mile Island Unit 2 (TMI - 2) reactor cavity (i.e., the annular gap between the pressure vessel and the biological shield) for nondestructive assessment of the fuel distribution. Two axial stringers were deployed in the annular gap with 17 SSTR dosimeters located on each stringer. SSTR experimental results reveal that neutron streaming, upward from the bottom of the reactor cavity region, dominates the observed neutron intensity. These absolute thermal neutron flux observations are consistent with the presence of a significant amount of fuel debris lying at the bottom of the reactor vessel. A conservative lower bound estimated from these SSTR data implies that at least 2 tonnes of fuel, which is roughly 4 fuel assemblies, is lying at the bottom of the vessel. The existence of significant neutron streaming also explains the high count rate observed with the source range monitors (SRMs) that are located in the TMI-2 reactor cavity.