Determination of critical mass and neutron flux distribution by means of physical models

The design of nuclear reactors, especially new reactors, requires experimental measurements in order to obtain accurate values of the pertinent parameters. In the present paper we present a new method for the preliminary determination of the critical mass of a reactor and the neutron flux distribution; this method is based on the use of physical models. In carrying out these experiments use is made of a model of the reactor which does not contain fissionable material. The working channels in the model are filled with a neutron absorber whose cross section simulates the absorption cross section for neutrons in the fissionable material. The production of fast fission neutrons is simulated by means of a neutron source which is moved along the channels. The distribution of thermal neutrons is measured by means of detectors which are sensitive to thermal neutrons. If the source strength and the absolute value of the neutron flux are known, it is possible to find the critical mass of the reactor.This method has been checked in a reactor with uranium hexafluoride. The value of the critical mass found experimentally was found to be in good agreement with the value obtained when the reactor was started up.The proposed method can also be useful in preliminary investigations of reactor designs, the choice of optimum lattice parameters, etc. The technique is extremely simple and does not require fissionable material or high neutron fluxes.     

Tory II-C Nuclear Instrumentation System

The design of the nuclear instrumentation system for the Pluto series of nuclear ramjet test reactors is an attempt to provide a very flexible nuclear sensing system that will be adequate for Tory II-C and following test reactors. The nuclear detectors will be exposed to the leakage neutron flux from the reactor during operation. Since the leakage flux is proportional to reactor power, the neutron detectors will give a measure of reactor power. A difficulty in providing nuclear instruments for this reactor is the uncertainty in the neutron energy spectrum of the leakage flux at the detectors. Since detector response varies with neutron energy, a large margin of flexibility is desirable. A difficulty which may be encountered is a significant shift in neutron energy spectrum at high power and temperature. This would make indicated nuclear power nonlinear with calorimetric power. A difficulty in insufficient instrument overlap was encountered with the Tory II-A experiment where a large margin of flexibility would have been useful. The detector placement for the Tory II-A experiment had the power range detectors in line with the reactor and main air pipe. At high air flows there was a much greater mass of air between the detectors and the reactor, allowing fewer neutrons to reach the detectors per unit reactor power. This is the reason for the power range detectors being placed off to the side of the Tory II-C test vehicle. Not all difficulties can be foreseen, but provision is made where possible to overcome them.     

Production of transuranic elements in the SM-2 and mir reactors

Conclusions The relative rates of accumulation of the individual isotopes of the transuranic elements (and hence also the isotopic compositions of these elements) may vary over a wide range in accordance with the conditions of irradiation of the initial materials. The effective cross sections for the capture of the neutrons by even-even and odd-even nuclei increase substantially as the proportion of resonance neutrons in the reactor spectrum incrases. Hence the irradiation of the original materials in the hard spectrum of the active zone of the SM-2 leads to the formation of elements with a high concentration of isotopes having an odd number of neutrons. This enables us to produce elements with sharply differing isotopic compositions, which in turn eases the study of the nuclear properties of individual isotopes.The successful combination of the high thermal-neutron flux in the trap of the SM-2 reactor, the hard neutron spectrum in the active zone of this reactor, and the large spaces available for irradiation in the MIR enables us to accumulate the desired isotopes under almost optimum conditions in every case.Translated from Atomnaya, Énergiya, Vol. 33, No. 4, pp. 815–819, October, 1972.     

Effects of Non-uniform Moderator Temperature on Core Power Distribution in Accelerator Driven Reactors

In a thermal reactor with moderators at different temperatures, a difference arises in the average speeds of thermalized neutrons between the high temperature part and the low temperature part of the moderator, and the non-uniformity of the average speed of thermalized neutrons may effect changes in the spatial dependence of the thermal neutron flux in a core. To investigate the thermal neutron flux in the case the average speed of thermalized neutrons is dependent on the position within a core, time-dependent two-group diffusion equations were applied. The influence of a nonuniform moderator temperature on the core power distribution was investigated about a graphite-moderated subcritical reactor driven by periodic injections of pulsed fast neutrons. The cylindrical reactor model by which a high temperature part of a core that has a spallation target at the center is enclosed by a low temperature part of a core was used. Changes in the core power distribution were calculated. It turned out that the momentary increases of a thermal power density caused by periodic injections of pulsed fast neutrons increase as the difference in the average speeds of thermalized neutrons in the high temperature part and the low temperature part of a core increases.     

Core Safety Characteristics of High Flux Safe Reactor (HFSR)

Abstract

Transient analyses are performed for graphite moderated helium-cooled high flux reactor to obtain the high flux safe reactor design. In order to promote the safety of the high flux reactor, the present design adopts the pebble bed reactor and its fuel technology. In the transient analyses, among the postulated off-normal events and accidents, the reactivity accident followed by a loss of helium forced circulation with system depressurization is found to be the severest potential event which may threaten the reactor safety from fission products release point of view. Several neutronic and thermal-hydraulic design parameters are indicated and exploited to promote the reactor safety. Neutronic and core thermal-hydralic models are proposed and used to simulate the reactor responses to the off-normal events and accidents. As the results of the transient analyses and accident simulation, safe and optimal design parameters are obtained which provide high thermal neutron flux with a desirable spectrum and large usable volume constrained by safety limitations.     

Implications of the differences in neutron spectra predicted for reactor pressure vessel walls by transport and the P1MG codes

Neutron-energy spectra were calculated for the interface between the vessel wall and cladding of the Army SM-1A Reactor pressure vessel using the transport theory code Program S and the diffusion code P1MG. Different sets of basic nuclear data and microscopic cross sections were used for the two calculations. Spectra were normalized to the same amount of activation in an iron, neutron flux detector. The transport code predicted a higher flux of neutrons in the energy groups between 6 and 10 MeV resulting in a lower overall intensity for the transport theory spectrum versus the P1MG spectrum. This was found to be consistent with the predictions of two transport codes versus the P1MG code for the PM-2A reactor vessel wall and for a simulated reactor vessel wall experiment. Such divergence of results for a given reactor using two different code analysis techniques raises important questions as to their usually unqualified acceptance and use for projecting the lifetime fluence for a reactor pressure vessel. Strong support is thus generated for establishment of one “standard” set of basic nuclear data from which all reactor physics analysts can draw to generate specific cross sections for reactor physics calculations, and for the writing of a new reactor physics spectrum code specifically for deep penetration analysis of reactor pressure vessel walls.     

Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum.This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG).The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra.     

Submersion criticality safety of fast spectrum space reactors: Potential spectral shift absorbers

Compact, fast spectrum, nuclear reactors are being considered to support NASA's future space exploration sometime in the next decade. In order to secure launch approval, these reactors should remain sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. In such an accident, the neutron spectrum in the reactor is thermalized, typically increasing reactivity, and potentially making the reactor supercritical. Incorporating “Spectral Shift Absorbers” (or SSAs), which have significantly higher absorption cross-sections for thermal versus fast neutrons, could offset the reactivity increase. It has always been the assertion that the worst-case submersion accident involves a fully flooded reactor; however, this work shows that, depending on the type and amount of SSA in the reactor, a submerged but unflooded reactor could be more reactive. A screening of the existing nuclear database for potential SSAs yielded 28 elements and nuclides, which are examined in detail as additives to a representative homogenous space reactor core by varying the SSA-to-U²³⁵ atom ratio. The effect of placing a thin coating of different SSA materials on the outside surface of the reactor core is also investigated. Nine SSAs (boron-10, cadmium, cadmium-113, samarium-149, europium-151, gadolinium, gadolinium-155, gadolinium-157, and iridium) are recommended for further consideration in actual space reactor designs.     

Measurement of neutron spectra in the thermal column of the reactor of a power station

Knowledge of neutron spectra In nuclear reactors allows comparison of various theories of the slowing down of neutrons with experiment, and also allows carrying out reactor calculations which are based on actual neutron distributions therein. In this paper is described a neutron intensity monochtomator Intended for the measurement of neutron spectra in the energy interval 0 to 0.5 ev.Results are given for measurements for neutron spectra in the thermal column of the reactor of an atomic power station. Discontinuities in the neutron flux were discovered at neutron velocities of 600, 1000 and 1650 m/sec; an analysis is given of the causes of discontinuities of the neutron flux; an evaluation is made of the inelastic scattering cross section for neutrons in graphite. By the method of least squares, the temperature of the neutron gas was found, being equal to 354 ° K at a graphite temperature of 304 ° K.In conclusion we consider it our duty to express gratitude to A. K. Krasin and B. G. Dubovskii for interest and help in the work and F. L. shapiro for valuable interpretation of previous results.     

The Void and Hole Effect in D2 Moderated Reactor

The reactivity cbange due to increase in the radius of empty hole was measured in a D₂O moderated reactor and some results differing from experiments with ZEEP were obtained. It can be concluded that the streaming in a hole is not so effective for reactivity. In measuring neutron flux in a void, a flat thermal neutron flux distribution was obtained and it has been concluded that the neutrons leaking through the empty hole or the void do not consist of thermal neutrons but fast neutrons for the most part. The experimental result of reactivity change due to the void location in the core indicates that the relation between the void location and the reactivity change is independent of the neutron flux distribution.     

Neutron irradiation studies on low density pan fiber based carbon/carbon composites

Carbon has been extensively used in nuclear reactors and there has been growing interest to develop carbon-based materials for high-temperature nuclear and fusion reactors. Carbon-carbon composite materials as against conventional graphite material are now being looked into as the promising materials for the high temperature reactor due their ability to have high thermal conductivity and high thermal resistance. Research on the development of such materials and their irradiation stability studies are scant. In the present investigations carbon-carbon composite has been developed using polyacrylonitrile (PAN) fiber. Two samples denoted as Sample-1 and Sample-2 have been prepared by impregnation using phenolic resin at pressure of 30 bar for time duration 10 h and 20 h respectively, and they have been irradiated by neutrons. The samples were irradiated in a flux of 10¹² n/cm²/s at temperature of 40 °C. The fluence was 2.52 × 10¹⁶ n/cm². These samples have been characterized by XRD and Raman spectroscopy before and after neutron irradiation. DSC studies have also been carried out to quantify the stored energy release behavior due to irradiation. The XRD analysis of the irradiated and unirradiated samples indicates that the irradiated samples show the tendency to get ordered structure, which was inferred from the Raman spectroscopy. The stored energy with respect to the fluence level was obtained from the DSC. The stored energy from these carbon composites is very less compared to irradiated graphite under ambient conditions.     

Attenuation of reactor thermal neutrons in a bulk shield of ordinary concrete

This work is concerned with the study of the distribution and attenuation of doses of thermal neutrons emitted directly from the core of ²³⁵U research reactor in ordinary concrete shields. In practice it is not possible to identify the reactor thermal neutrons in the emitted continous neutron spectrum, therefore, measurements were carried out by using a direct and cadmium filtered beam of reactor neutrons. All measurements were performed using Li₂B₄O₇:Mn thermoluminescent dosimeters.The data obtained were analysed and the dose distributions of reactor thermal neutrons were evaluated. A group of isodose curves were constructed which give directly the shape and thickness of the shield required to attenuate the intensity of doses of reactor thermal neutrons to specific values. In addition, the thermal neutron relaxation lengths in ordinary concrete were derived for disc collimated beam and infinite plane monodirectional sources.     

Computational-experimental studies of the neutronics of the SM reactor with different variants of the neutron-trap arrangement

The results of experimental studies of the neutronics of the high-flux SM reactor with different arrangements of the neutron trap are presented. The MCU series of high-precision computer programs implementing the Monte Carlo method is used for computations. Experimental data on reactivity effects, the effectiveness of safety and control rods, and the coefficients of nonuniformity of energy release in the core have been obtained in experiments on a critical assembly – a physical model of the SM reactor – and directly in experiments in the reactor. The error is 4.2–10% in determining the reactivity parameters and 5–10% for the relative energy release in the fuel elements. Information on the neutron field formed in the volume of the neutron trap has been obtained for two arrangements of the beryllium and water moderators. The differential and integral energy spectra of the neutrons in the energy interval from 0.5 eV to 20 MeV are obtained for three points inside the trap (external, central series, center). The flux density of thermal, superthemal, and fast neutrons are determined.     

Nuclear burning wave in fast reactor with mixed Th-U fuel

The possibility of creating a self-sustained regime of a running nuclear burning wave in the critical fast reactor with the mixed Th-U fuel is demonstrated. The calculations were performed in the deterministic approach based on solving the non-stationary multi-group diffusion equation of neutron transport together with the set of equations of the fuel component burn-up and the nuclear kinetics of precursor nuclei of delayed neutrons. The presence of the constructional material Fe and the coolant (the Pb-Bi eutectic) in the reactor composition is taken into account. The calculation results of the space-time evolution of neutron flux and fuel component concentrations are presented for different values of the Th-U ratio in the fuel. The calculations show the remarkable stability of the nuclear burning wave regime against neutron flux distortions in the reactor, which is a result of the negative feedback on reactivity inherent to this regime. This is one of the most important features of the reactor of this type, which ensures its intrinsic safety.     

Investigations on the migration mode method (MMM) for reactor calculations

Current calculation codes for reactor analysis are based on the multi-group method to evaluate energy distribution of neutron flux. Usually a two energy group diffusion equation is adopted. This choice is adequate for PWRs associated to cross sections libraries tabulated versus fuel exposure and other state parameters as moderator density, fuel temperature, boron concentration. An improvement of this approach is represented by the migration mode method (MMM) by which the neutron spectrum is expanded in terms of base functions corresponding to the different modes of migration of the neutrons in the energy dimension. For a thermal reactor, three such functions may be easily identified: the Maxwellian distribution of the neutrons at thermal equilibrium with the moderator, the 1/E slowing down distribution (corrected to take into account resonance absorption effects) and the fission neutron spectrum. The (space-dependent) coefficients of the expansion are calculated by solving a differential equation which results having a structure similar to the one relevant to multi-group theory. The method can therefore be easily implemented adopting existing diffusion theory codes.     

Production of weighable quantities of Tc99 from molybdenum irradiated with neutrons

Various methods of separating technetium from molybdenum anhydride irradiated with thermal neutrons in a nuclear reactor have been investigated. A method has been devised and tested In laboratory conditions for the concentration and separation of technetium, based on the co-precipitation of technetium with difficultly soluble phosphates and on Chromatographic purification. Milligram quantities of Tc⁹⁹ have been obtained by this method. The separated technetium has been identified by means of spectrum analysis, and the absolute activity and maximum energy of its ß -radiations have also been measured. Some chemical properties of technetium have been studied.Presented on March 9, 1957 at the All-Union Radiochemistry Conference in Leningrad.     

Nuclear Analysis of an ITER Blanket Module

ITER blanket system is the reactor’s plasma-facing component, it is mainly devoted to provide the thermal and nuclear shielding of the Vacuum Vessel and external ITER components, being intended also to act as plasma limiter. It consists of 440 individual modules which are located in the inboard, upper and outboard regions of the reactor. In this paper attention has been focused on to a single outboard blanket module located in the equatorial zone, whose nuclear response under irradiation has been investigated following a numerical approach based on the Monte Carlo method and adopting the MCNP5 code. The main features of this blanket module nuclear behaviour have been determined, paying particular attention to energy and spatial distribution of the neutron flux and deposited nuclear power together with the spatial distribution of its volumetric density. Moreover, the neutronic damage of the structural material has also been investigated through the evaluation of displacement per atom and helium and hydrogen production rates. Finally, an activation analysis has been performed with FISPACT inventory code using, as input, the evaluated neutron spectrum to assess the module specific activity and contact dose rate after irradiation under a specific operating scenario.     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

Radiation Field for Fission Neutron Enhanced Boron Neutron Capture Therapy Employing Fast Neutron Source Reactor “YAYOI”

Boron Neutron Capture Therapy (BNCT) of a localized tumor needs a sufficient thermal neutron flux at the tumor. A surgical operation including ennucleation of the main part of tumor is required for the case of thermal neutron beam from a thermal reactor because of the rapid decrease of the neutron flux in the tissue. Intermediate neutrons with little fast neutron component are only produced by a specifically designed reactor which awaits to be build.

In the present paper, a positive use of fast neutron beams in addition to BNCT is proposed for treatment of some kind of localized tumors employing a fission fast neutrons from a fast neutron source reactor “YAYOI” of University of Tokyo which is licenced as such. Dose distributions in a water phantom located at a proposed position for two collimator cases were measured and its availability was confirmed as a possible port for therapy.