Basic study on neutronics of future neutron source based on accelerator driven subcritical reactor concept in Kyoto University Research Reactor Institute (KURRI)

A basic study on the nuclear characteristics in the accelerator driven subcritical reactor (ADSR) was performed through a series of neutronics design calculations and reactor physics experiments. Calculations were executed mainly by the MCNPX code, and experiments were performed at the Kyoto University Critical Assembly (KUCA). Some nuclear features of the research reactor type ADSR were revealed through the present study. The following facts were found: 1) Further studies are necessary concerning the nuclear data in the high energy region and the generated neutrons through the spallation reactions especially by the light nuclei and the lower energy protons. 2) The adjustment of subcriticality by the control rod significantly affects the reactor power of ADSR because of the distortion in the neutron flux distribution caused by the control rod insertion. 3) An accurate calculation is essential to evaluate the neutron multiplication in the ADSR. 4) The neutronics behavior after a pulse injection can be approximately simulated by the calculation.     

Preliminary Experiments on Accelerator-Driven Subcritical Reactor with Pulsed Neutron Generator in Kyoto University Critical Assembly

A series of preliminary experiments on an accelerator-driven subcritical reactor (ADSR) with 14 MeV neutrons were conducted at Kyoto University Critical Assembly (KUCA) with the prospect of establishing a new neutron source for research. A critical assembly of a solid-moderated and -reflected core was combined with a Cockcroft-Walton-type accelerator. A neutron shield and a beam duct were installed in the reflector region for directing as large a number as possible of the high-energy 14MeV neutrons generated by deuteron-tritium (D-T) reactions to the fuel region, since the tritium target is located outside the core. And then, neutrons (14MeV) were injected into a subcritical system through a polyethylene reflector. The objectives of this paper are to investigate the neutron design accuracy of the ADSR with 14MeV neutrons and to examine experimentally the neutronic properties of the ADSR with 14MeV neutrons at KUCA. The reaction rate distribution and the neutron spectrum were measured by the foil activation method for investigating the neutronic properties of the ADSR with 14 MeV neutrons. The eigenvalue and fixed-source calculations were executed using a continuous-energy Monte Carlo calculation code MCNP-4C3 with ENDF/B-VI.2 for the subcriticality and the reaction rate distribution, respectively; the unfolding calculation was done using the SAND-II code coupled with JENDL Activation Cross Section File 96 for the neutron spectrum. The values of the calculated subcriticality and the reaction rate distribution were in good agreement with those of the experiments. The results of the experiments and the calculations demonstrated that the installation of the neutron shield and the beam duct was experimentally valid and that the MCNP-4C3 calculations were accurately carried out for analyzing the neutronic properties of the ADSR with 14MeV neutrons at KUCA.     

Neutronics analysis of a stacked structure for a subcritical system with LEU solution driven by a D-T neutron source for 99Mo production

The utilization of neutrons markedly affects the medical isotope yield of a subcritical system driven by an external D-T neutron source.The general methods to improve the utilization of neutrons include moderating,multiplying,and reflecting neutrons,which ignores the use of neutrons that backscatter to the source direction.In this study,a stacked structure was formed by assembling the multiplier and the low-enriched uranium solution to enable the full use of neutrons that backscatter to the source direction and further improve the utilization of neutrons.A model based on SuperMC was used to evaluate the neu-tronics and safety behavior of the subcritical system,such as the neutron effective multiplication factor,neutron energy spectrum,medical isotope yield,and heat deposi-tion.Based on the calculation results,when the intensity of the neutron source was 5×1013 n/s,the optimized design with a stacked structure could increase the yield of 99Mo to 182 Ci/day,which is approximately 16％higher than that obtained with a single-layer structure.The inlet H2O coolant velocity of 1.0 m/s and initial temperature of 20℃were also found to be sufficient to prevent boiling of the fuel solution.     

Comparison of calculated and measured reaction rates obtained through foil activation in the subcritical dual spectrum facility YALINA-Booster

Reaction rates were measured by the foil activation technique to obtain neutron spectrum information in a subcritical core driven by an external neutron source. The experimental results are compared with Monte Carlo calculations in order to examine the capability of the Monte Carlo code MCNP together with ENDFB-6.8, JEFF-3.1.1 and CENDL-3.1 neutron cross section libraries to predict the neutron spectrum dependent reaction rates correctly in a subcritical core. The focus lies on fast neutrons. A discrepancy is found in the calculated-to-experimental values of the reaction rates and an inaccurate cross section is identified in CENDL-3.1.     

Study on kinetics of fast subcritical assembly—Contribution of delayed neutrons to the transition after a reactivity insertion in a subcritical system—

This study is an investigation of the effect of the delay neutron on the kinetics in the subcritical system. And, it proposes a method necessary for the kinetics code development that uses the Monte Carlo (MC) computation.

It is generally difficult to analyze three dimensional space and time dependent kinetics by using a MC method. It is because the sampling of the neutron in a region becomes difficult when conditions of the region changes with time. In this study, we consider about the effect of delayed neutron in the kinetics of ADS. The behavior of neutrons is considered spontaneous in this system. It means a neutron is absorbed or leaks in a short period, while the conditions of region do not change. Therefore they are treated by steady state calculation. On the other hand the densities of delayed neutron precursors changes slowly, and the conditions of region change. In the concept of developed MC method, the neutrons are calculated by using steady state equation at each time point, and the delayed neutron precursors are calculated by using time dependent equation. We tried to inspect the accuracy of this method by using a point equation. We obtained strict solution Φ* as a reference solution, Φ1 as a solution by the present method, and Φ2 as the solution where both neutrons and delayed neutron precursors are treated by using static equations. The obtained results show a good agreement between Φ1 and Φ*, though the Φ2 agrees with Φ* poorly in all cases. Especially, we showed that this technique was effective from the reactivity change by ADS, and the relation of a delayed neutron. Finally, the effect of the delay neutron on the beam trip in the neutron source for the drive was examined by using the technique of Φ2.     

Multipurpose electron accelerator driven electronuclear system based on a subcritical cascade reactor

An electronuclear system consisting of an electron accelerator, a neutron-producing target, and dual-zone subcritical blanket with fast and thermal neutron spectra is proposed. Some general mechanisms of cascade multiplication of neutrons in a dual-zone subcritical blanket are examined. The results of calculations of the electron-photon-neutron interactions of an electron beam with the target material and of the neutron-physical and heat-engineering characteristics of the system are presented. It is shown that a cascade neutron amplification factor of 2.5–3 can be obtained with system subcriticality 2%. The power of the system reaches 50 MW with electron beam power 4 MW. __________ Translated from Atomnaya énergiya, Vol. 102, No. 2, pp. 92–98, February, 2007.     

Experimental study of neutron counting in a zero-power reactor driven by a neutron source inherent in highly enriched uranium fuels

Even a zero-power reactor core containing highly enriched uranium has a weak neutron source inherent in uranium 235, and consequently, a neutron counter placed closely to the core without external neutron source registers a certain counting rate. The study of the counting is very important for zero-power reactor physics experiments with a high precision. In this experimental study, first, at a shutdown state of the UTR-Kinki reactor without start-up neutron source, a pulse height distribution of output signals from a neutron proportional counter was measured to confirm that these signals resulted from neutron detections. At several subcritical states of the UTR, then, the Feynman-α analysis was carried out to confirm that the neutrons detected by the counter must be fission neutrons multiplied by fission chain reactions. The correlation amplitude measured in the Feynman-α analysis was much higher than that measured in a previous drive by start-up source. Further, it was also confirmed that the subcriticality dependence of neutron counting rate followed the source multiplication formula. This feature indicated that the one-point model was very successful in the subcritical range including the shutdown state.     

Power spectral analysis for a thermal subcritical reactor system driven by a pulsed 14 MeV neutron source

A series of power spectral analyses for a thermal subcritical reactor system driven by a pulsed 14 MeV neutron source was carried out at Kyoto University Critical Assembly (KUCA), to determine the prompt-neutron decay constant of the accelerator-driven system (ADS). The cross-power spectral density between time-sequence signal data of two neutron detectors was composed of a familiar continuous reactor noise component and many delta-function-like peaks at the integral multiple of pulse repetition frequency. The prompt-neutron decay constant inferred from the reactor noise component of the cross-power spectral density was consistent with that obtained by a pulsed neutron experiment. However, the reactor noise component of the auto-power spectral density of each detector was hidden by a white chamber noise in the higher-frequency range and this feature resulted in a considerable underestimation of the decay constant. For several runs with a low pulse-repetition frequency, furthermore, we attempted to infer the decay constant from point data of the delta-function-like peaks. The analysis for a run under a slightly subcritical state resulted in the consistent decay constant; however, those for other runs under significantly subcritical states underestimated the decay constant. Considering the contribution of a spatially higher mode to the point data, the above underestimation was solved to obtain the consistent decay constant. While the Feynman-α formula for a pulsed neutron source is too complicated to be fitted directly to variance-to-mean ratio data, the present analysis on frequency domain is much simpler and the conventional formula based on the first-order reactor transfer function is available for fitting to power spectral density data.     

Preparation and verification of mixed high-energy neutron crosssection library for ADS

An accelerator-driven subcritical system(ADS)is driven by an external spallation neutron source, which is generated from a heavy metal spallation target to maintain stable operation of the subcritical core, where the energy of the spallation neutrons can reach several hundred megaelectron volts. However, the upper neutron energy limit of nuclear cross-section databases, which are widely used in critical reactor physics calculations, is generally 20 MeV.This is not suitable for simulating the transport of highenergy spallation neutrons in the ADS. We combine the Japanese JENDL-4.0/HE high-energy evaluation database and the ADS-HE and ADS 2.0 libraries from the International Atomic Energy Agency and process all the data files for nuclides with energies greater than 20 MeV. We use the continuous pointwise cross-section program NJOY2016 to generate the ACE-formatted cross-section data library IMPC-ADS at multiple temperature points. Using the IMPC-ADS library, we calculate 10 critical benchmarks of the International Criticality Safety Benchmark Evaluation Project manual, the 14-MeV fixed-source problem of the Godiva sphere, and the neutron flux of the ADS subcritical core by MCNPX. To verify the correctness of the IMPCADS, the results were compared with those calculated using the ENDF/B-VII.0 library. The results showed thatthe IMPC-ADS is reliable in effective multiplication factor and neutron flux calculations, and it can be applied to physical analysis of the ADS subcritical reactor core.     

Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.     

Neutron damage calculations in Cu,Nb and Au to 32 MeV: Application to sputtering and deuteron-breakup neutron sources

Primary recoil distributions and specific damage energies have been computed for high energy deuteron-breakup neutrons in Cu, Nb and Au. The calculations are based on theoretical neutron cross sections and consider in particular a d-Be spectrum broadly peaked at 15 MeV with some neutrons above 30 MeV. The theoretical results are similar to corresponding calculations for monoenergetic 15-MeV neutrons and are in good agreement with range measurements of (n, 2n) recoils generated by high energy d-Be neutrons in Nb and Au. The calculations are also consistent with recent d-Be neutron sputtering experiments in Nb and Au and demonstrate the usefulness of deuteron-breakup neutron sources for simulating fusion neutron effects.     

Study of the spectra and doses created in the iron-water shielding of a monoenergetic neutron source

Many-group calculations were made for the penetration of neutrons, emitted from monoenergetic sources, through water, iron, andwater-iron systems of finite dimensions; the results of these calculations are presented. The neutron spectra resulting from the passage of such neutrons through water and iron shielding layers were calculated on the twenty-group diffusion-transport approximation, Detailed attention was paid to the high-energy part of the spectrum; certain peculiarities in neutron migration and moderation processes in shielding of the type in question were elucidated. Dose curves D(r) were plotted for neutrons of various energies.By using the superposition principle, the results enable the neutron spectrum to be determined for sources having any arbitrary spectrum.Translated from Atomnaya Énergiya, Vol. 21, No. 1, pp. 27–35, July, 1966.     

Feynman-Y function for a subcritical assembly with intrinsic spontaneous fissions driven by external pulsed sources

Stochastic neutron transport theory is applied to the derivation of the Feynman-Y function for subcritical assemblies when external pulsed sources are used. We obtain a general relationship between the probability generating functions of the kernel and the source considering the contribution to the detector statistics of both the pulsed source and the intrinsic neutron source. An expansion in α-eigenvalues is derived for the final solution, which permits to take into account the effect of higher harmonics in subcritical systems. In addition, numerical calculations have been done for a proposed model problem to understand the applicability of the method studied.     

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.     

Measurements of neutron induced activation of concrete at 64.5 MeV

The present paper presents the measurement of neutron induced activations on concrete using the 64.5 MeV quasimonoenergetic neutrons produced at the intense ⁷Li(p, n) neutron source at Cyclotron and Radioisotope Center, Tohoku Univeristy (CYRIC). The data were corrected for the effect of continuous neutrons in the source. The neutron energy, neutron yields and the spectrum of continuous neutrons were confirmed with the neutron time-of-flight method and the neutron activation measurement of the ²⁰⁹Bi(n, Xn) reactions having various threshold energy values. The nuclides produced by thermalized source neutrons are negligible. New data were obtained for concrete activation.     

Feynman-α and Rossi-α analyses for a subcritical reactor system driven by a pulsed spallation neutron source in Kyoto University Critical Assembly

ABSTRACT

For a subcritical reactor system driven by a periodically pulsed spallation neutron source in Kyoto University Critical Assembly (KUCA), the Feynman-α and the Rossi-α neutron correlation analyses were carried out to determine the prompt-neutron decay constant and quantitatively to confirm a non-Poisson characteristics of the neutron source. In these correlation analyses, a non-negligible contribution of delayed neutrons and a non-Poisson character of the source were considered, and each pulse was assumed to be a delta function. When a neutron counter was placed closely to the reactor core, the prompt-neutron decay constant determined from the present Feynman-α analysis well agreed with that done from a previous analysis for the same subcritical system driven by an inherent neutron source. However, the decay constant determined from the present Rossi-α analysis was in poor agreement with that done from the above previous analysis. This disagreement originated from an inevitable excitation of a higher mode. In the Rossi-α counting probability distribution, the excitation deformed a sharp cusp arising from the delta function to a smooth convex shape. When the data around the convex top were masked for least-squares fitting of the present Rossi-α formula, the disagreement could be successfully resolved. Compared with the previous Feynman-α and Rossi-α analyses under the Poisson inherent source, the non-Poisson spallation source definitely enhanced the respective prompt-neutron correlation amplitudes. The enhancement rate increased with an increase in subcriticality. Moreover, the Degweker’s factor (m ₂-m ₁ ²)/m ₁ ² of 0.067 ± 0.011, which indicated a non-Poisson character of the present spallation source, could be determined from the present correlation analysis and the non-zero value of the factor convinced us that the present source had a different statistical distribution from the Poisson.     

Reaction rate analyses of accelerator-driven system experiments with 100 MeV protons at Kyoto University Critical Assembly

At the Kyoto University Critical Assembly, a series of reaction rate experiments is conducted on the accelerator-driven system (ADS) with spallation neutrons generated by the combined use of 100 MeV protons and a lead–bismuth target in the subcritical state. The reaction rates are measured by the foil activation method to obtain neutron spectrum information on ADS. Numerical calculations are performed with MCNP6.1 and JENDL/HE-2007 for high-energy protons and spallation process, JENDL-4.0 for transport and JENDL/D-99 for reaction rates. That the reaction rates depend on subcriticality is revealed by the accuracy of the C/E (calculation/experiment) values. Nonetheless, the accuracy of the reaction rates at high-energy thresholds remains an important issue in the fixed-source calculations. From reaction rate analyses, the indium ratio is newly defined as another spectrum index with the combined use of ¹¹⁵In(n, γ)^116mIn and ¹¹⁵In(n, n′)^115mIn reaction rates, and considered useful in examining the neutron spectrum information on ADS with spallation neutrons.     

Study of the neutron multiplication effect in an active neutron method

The neutron multiplication effect appears when an item contains large amounts of nuclear material. The neutron multiplication effect in this paper means the effect of subsequent fission reactions which are caused by fission neutrons produced by interrogation neutrons from a neutron generator. The previous active neutron method could not distinguish between first-fission and subsequent-fission neutrons and might overestimate the amount of nuclear material. However, the neutron multiplication effect in the active neutron method has not been adequately investigated. We discuss the evaluation method of the multiplication effect in the fast neutron direct interrogation method, one of the active neutron methods, using simulations with the Monte Carlo code MVP and experiments involving uranium waste drums. The first-generation neutrons from an external neutron source generate fission neutrons called second-generation neutrons, the second-generation neutrons generate third-generation neutrons, and so on. This study supposes that the neutron multiplication effect is mainly caused by the third-generation neutrons under the condition that the fourth-generation neutrons are much fewer. This paper proposes a correction method for the neutron multiplication effect in the measured data.     

Source reactivity as an extra kinetic characteristic of coupled-source subcritical systems

The objective of the present work is to study the kinetic (i.e., in absence of in-core feedbacks) response of the subcritical system to variation of the efficiency of the external neutron source. The particular class of the subcritical system with the intensity of the external source being intrinsically dependent on neutron production in the core (so-called coupled hybrid systems) is considered. The accelerator coupled system is taken as example, although this analysis may be expanded to other types of the coupled subcritical systems. Within the framework of a simple mathematical model of coupled system, an interpretation of the external coupled source as supplementary group of delayed neutrons is given. An auxiliary quantity – ‘source reactivity’ is introduced for convenience and a modified inhour equation for coupled systems is deduced. Analytical solution of the modified inhour equation is obtained in approximation of one group of delayed neutrons. The principal conclusion resulting from this analysis is as follows: the response of the coupled system to ‘source reactivity’ variation is intrinsically different from the response to core reactivity variation. Namely, there is no equivalent of prompt criticality (accompanied by drastic decrease of the reactor period) in the case of ‘source reactivity’ variation.     

Effects of Spatial Harmonics on Power Distribution in Accelerator Driven Subcritical Cores

A method for solving the time-energy-dependent diffusion equation was devised to appreciate the effect of spatial harmonics on the power distribution in a subcritical graphite-moderated core with a pulsed fast neutron source. An instantaneous power peaking factor (IPF) was calculated using the thermal neutron flux thus obtained, to characterize the power distribution in the core. In case of a 50-cycles-per-second injections of fast neutron pulses of 3 ms pulse width, it was indicated that the value of IPF increased by 4.1 times with the decrease in the multiplication factor from 0.9951 to 0.9762, accompanying fuel burnup. To appreciate the dependence of the core power distribution on the spectrum hardening of the thermal neutron flux caused by injections of pulsed fast neutrons, comparison was made between the value of IPF and that derived from the time-dependent-one-group diffusion theory where the spectrum hardening of the thermal neutron flux was ignored. It was indicated that the disregard of the spectrum hardening of the thermal neutron flux resulted in the low estimate of IPF that changed from ?1% to ?14% with the decrease in the pulse width of pulsed fast neutrons from 3 ms to 0.15 ms.