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.     

Materials researches at the Paul Scherrer Institute for developing high power spallation targets

For the R&D of high power spallation targets, one of the key issues is understanding the behavior of structural materials in the severe irradiation environments in spallation targets. At PSI, several experiments have been conducted using the targets of the Swiss spallation neutron source (SINQ) for studying radiation damage effects induced by high energy protons and spallation neutrons. As well, experiments have been performed to investigate liquid lead-bismuth eutectic (LBE) corrosion and embrittlement effects on T91 steel under irradiation with 72 MeV protons. In this paper, an overview will be given showing a selection of results from these experiments, which include the mechanical properties and microstructure of ferritic/maretensitic (FM) steels (T91, F82H, Optifer etc.) and austenitic steels (EC316LN, SS 316L, JPCA etc.) irradiated to doses higher than ever attained by irradiation in a spallation environment, and the behaviors of T91 irradiated with 72 MeV protons in contact with flowing LBE.     

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.     

Uncertainty quantification of criticality in solid-moderated and -reflected cores at Kyoto University Critical Assembly

Uncertainty quantification is conducted for the criticality of excess reactivity and control rod worth obtained at the Kyoto University Critical Assembly (KUCA). By combining SRAC2006 and MARBLE code systems, the sensitivity coefficients of the cross sections for aluminum-27 (²⁷Al) comprising mainly of core components are large in the solid-moderated and -reflected cores (A cores) at KUCA. Also, the uncertainty is dominant in the uranium-235 isotope by the covariance data of JENDL-4.0, and a quantitative value is about 150 pcm induced by the JENDL-4.0 data library in the KUCA A cores, whereas the covariance data of ²⁷Al are not prepared in JENDL-4.0. Moreover, the effect of decreasing uncertainty is obtained by applying the cross-sectional adjustment method to the uncertainty analyses. From the results, a series of uncertainty quantifications is expected to clarify the uncertainty of sub-criticality in accelerator-driven system experiments with spallation neutrons in the KUCA A cores.     

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.     

Steady-state neutronic analysis of converting the UK CONSORT reactor for ADS experiments

CONSORT is the UK’s last remaining civilian research reactor, and its present core is soon to be removed. This study examines the feasibility of re-using the reactor facility for accelerator-driven systems research by replacing the fuel and installing a spallation neutron target driven by an external proton accelerator. MCNP5/MCNPX were used to model alternative, high-density fuels and their coupling to the neutrons generated by 230 MeV protons from a cyclotron striking a solid tungsten spallation target side-on to the core. Low-enriched U₃Si₂ and U–9Mo were considered as candidates, with only U–9Mo found to be feasible in the compact core; fuel element size and arrangement were kept the same as the original core layout to minimise thermal hydraulic and other changes. Reactor thermal power up to 2.5 kW is predicted for a k_eff of 0.995, large enough to carry out reactor kinetic experiments.     

Optimal ion beam,target type and size for accelerator driven systems: Implications to the associated accelerator power

In interactions of different energetic ions with extended targets hydrogen isotopes are the most effective projectiles for the production of spallation neutrons. It is shown that for every target material and incident ion type and energy there is an optimal target size which results in the escape of a maximum number of spallation neutrons from the target. Calculations show that in an ADS, combination of a beam of 1.5 GeV deuteron projectiles and a uranium target results in the highest neutron production rate and therefore highest energy gain. For fast 1.5 GeV d + ²³⁸U ADS with lead or lead–bismuth eutectic moderator, the required ion beam current is only 38% of that for 1 GeV proton projectiles on lead target. It is shown that for a modular ADS with uranium target and output power of 550 MW_th a 1.5 GeV deuteron beam of current 1.8 mA is required, which is easily achievable with today’s technology. For an ADS with k_eff = 0.98 and output power of 2.2 GW_th, the required beam currents for (a) 1 GeV p + Pb and (b) 1.5 GeV d + U systems are 18.5 and 7.1 mA, respectively.     

MgO single-crystal as an efficient thermal neutron filter

A formula is given which, for neutron energies in the range 10⁻⁴ < E < 10 eV, permits calculation of the nuclear capture, thermal diffuse and Bragg scattering cross-sections as a function of magnesium oxide (MgO) temperature and crystal parameters. Computer program has been developed to calculate the total neutron cross-section and transmission through mono-crystalline MgO. The calculated neutron transmission and effective attenuation coefficient values for MgO-single crystal at different temperatures are compared with measured ones. An overall agreement is indicated between the formula fits and experimental data. A feasibility study for the use of MgO-single crystal is discussed in terms of the optimum MgO-single crystal thickness, mosaic spread, temperature and cutting plane for efficient transmission of thermal-reactor neutrons, and also for rejection of the accompanying fast neutrons and gamma rays.     

Experimental analysis for neutron multiplication by using reaction rate distribution in accelerator-driven system

Basic experiments are carried out in the Kyoto University Critical Assembly (KUCA) to study the nuclear characteristics of the accelerator-driven systems (ADS) and to establish a new neutron source for research. A method for measuring the neutron multiplication in the subcritical system is proposed by utilizing the reaction rate distribution obtained by the optical fiber detector. A comparison between the measured and calculated neutron multiplication shows agreement within a relative difference of 5% in most cases, through introducing certain correction factors.     

The materials test station: A fast-spectrum irradiation facility

The United States Department of Energy is developing technologies needed to reduce the quantity of high-level nuclear waste bound for deep geologic disposal. Central to this mission is the development of high burn-up fuel with significant inclusion of plutonium and minor actinides. Different fuel forms (e.g., nitrides, oxides, and metal matrix) and composition are under study. The success of these cannot be judged until they have been irradiated and tested in a prototypic fast neutron spectrum environment. In 2005, the US Congress authorized funding for the design of the materials test station (MTS) to perform candidate fuels and materials irradiations in a neutron spectrum similar to a fast reactor spectrum. The MTS will use a 1-MW proton beam to generate neutrons through spallation reactions. The peak neutron flux in the irradiation region will exceed 1.2 × 10¹⁹ n m⁻² s⁻¹ and the fast neutron fluence will reach 2 × 10²⁶ n m⁻² per year of operation. Site preparation and test station fabrication are expected to take four years.     

Preliminary physics study of the Lead–Bismuth Eutectic spallation target for China Initiative Accelerator-Driven System

The Lead-Bismuth Eutectic(LBE) spallation target has been considered as one of the two alternatives for the spallation target for China Initiative Accelerator-Driven System.This paper reports the preliminary study on physical feasibility of a U-type LBE target with window.The simulation results based on Monte Carlo transport code MCNPX indicate that the spallation neutron yield is about 2.5 per proton.The maximum spallation neutron flux is observed at about 3 cm below the lowest part of the window.When the LBE target is coupled with the reactor,the reactor neutrons from the fission reaction increased the neutron field significantly.The energy deposition of highenergy protons is the main heat source;the spallation neutrons and reactor neutrons contribute only a small fraction.The maximum energy deposition in the LBE is about 590 W/cm~3 and that in the target window is about319 W/cm~3.To estimate the lifetime of the target window,we have calculated the radiation damages.The maximum displacement production rate in the target window is about10 dpa/FPY.The hydrogen and helium production rates generated during normal operation were also evaluated.By analyzing the residual nucleus in the target during the steady operation,we estimated the accumulated quantities of the extreme radioactivity toxicant ~(210)Po in the LBE target loop.The results would be helpful for the evaluation of the target behavior and will be beneficial to the optimization of the target design work of the experimental facilities.     

Heat deposition in thick targets due to interaction of high energy protons and thermal hydraulics analysis

Heat deposition inside thick targets due to interaction of high energy protons (E_p ∼ GeV) has been estimated using an improved version of the Monte Carlo simulation code CASCADE.04.h. The results are compared with the available experimental data for thick targets of Be, Al, Fe, Cu, Pb and Bi at proton energies of 0.8 GeV, 1.0 GeV and 1.2 GeV. A more continuous heat deposition approach which has been adopted in CASCADE.04.h yields results which are in better agreement with the experimental data as compared to the ones from the earlier version of CASCADE.04. The results are also compared with the predictions of the FLUKA Monte Carlo code. Both CASCADE.04.h and FLUKA predictions are nearly similar for heavy targets and both agree with the experimental measurements. However, they do have differences in predictions for lighter targets where measurements also differ from the predictions. It is observed that the maximum heat loss in thick targets occurs at the beginning of the target due to increasing nuclear reaction contributions. This aspect is crucial in designing the window of a spallation neutron target employed in an accelerator driven sub-critical system (ADS) as this is the first material to be traversed by the proton beam and is subjected to the maximum temperature gradient. Optimization of the target-window parameters requires a careful estimation of heat deposition in the window region and this has been demonstrated through thermal hydraulic studies related to the design of a realistic lead bismuth eutectic (LBE) spallation neutron target for an ADS system.     

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.     

Evaluation of the integral quantities for incident fusion source neutrons in infinite medium

Integral neutronic quantities have been evaluated for incident 14.1 MeV (D,T) and 2.45 MeV (D,D) fusion source neutrons in infinite medium for the most important materials in fusion technology, namely lithium, beryllium, lead and thorium. The study covers the calculation of the integral tritium breeding ratio, the ²³³U breeding rate (when applicable), the neutron multiplication ratio through (n,x)-and fission-(when applicable) reactions, and the heat release in those mixtures which are composed when first thorium is mixed with natural lithium or ⁶Li for a volume fraction from 0% to 100% and then the variable thorium-lithium composition is mixed with Be or Pb for a volume fraction from 0% to 100%.     

Cross sections for neutron-induced reactions up to 1.6 GeV for target elements relevant for cosmochemical, geochemical, and technological applications

We determined neutron excitation functions from the respective reaction thresholds up to 1.6 GeV for almost 100 target-product combinations relevant for cosmochemical, geochemical, and technological applications. We started with thick target production rates that have been obtained by irradiating iron and stone spheres with protons having energies between 600 MeV and 1.6 GeV. From the particle spectra of primary protons, secondary protons, and secondary neutrons, and the usually well known cross sections for the proton-induced reactions we calculated the production rates only due to protons. By subtracting these data from the measured total production rates we obtained production rates only due to secondary neutrons. With the modelled neutron spectra, guess functions calculated using nuclear model codes, and sophisticated energy-dependent deconvolution procedures we were able to determine almost 100 neutron excitation functions with their uncertainties. With the thus obtained neutron cross sections we are able to describe the experimental production rates in the thick target experiments, meteorites, the lunar surface, and terrestrial surface samples usually within the uncertainties, i.e., to within 10-15%. The adjusted neutron cross sections (a posteriori) are compared to results from the theoretical nuclear model codes INCL4.5/ABLA07 and TALYS. The TALYS code usually describes the a posteriori data reasonably well, i.e., mostly within a factor of a few. The quality of the INCL4.5 + ABLA07 predictions depends on the reaction type and increases with increasing number of ejectiles, i.e., increasing target-product mass difference. The neutron cross section database, though successful in quantifying production rates in terrestrial and extraterrestrial matter, presents by no means a final step and experiments with quasi-monoenergetic neutrons are needed.     

Sensitivity of fuel failure detection by the delayed neutron measurement in the primary cooling circuit of HANARO

The sensitivity of the fuel failure detection system based on the delayed neutron measurement in the primary cooling circuit of a research reactor, HANARO is investigated. The neutrons around the primary cooling pipe during normal operation of HANARO are measured with BF₃ detector, and their count rate is 900 cps. They are regarded as photoneutrons due to the high energy gamma-rays from N-16 and delayed neutrons from the fission of the uranium contaminated on the fuel surface. The contribution of each neutron source is analyzed by measuring the changes of the neutron counts before and after the abrupt shutdown of reactor. In order to estimate the sensitivity of the fuel failure detection, the neutron count rate of BF₃ detector is predicted by Monte Carlo calculation. The generation, transportation and detection of the photoneutrons and the delayed neutrons are simulated for the geometry similar to the experiments. From the calculations and experiments, it is ascertained that the photoneutron contribution to the total count rate is about 20–30%, and that the delayed neutron count rate is expected to about 720 cps. The fission rate in the flow tube of the reactor core by the surface contamination is obtained from the deduced delayed neutron count rate, and it is estimated to 1.66 × 10⁵ fissions/cm³ s. From the MCNP calculation, it is confirmed that this fission rate can originate from the contaminated uranium of 120 μg, which is about 13% of the maximum allowable surface contamination on the fuel surface. The sensitivity of U-235 mass detection by the delayed neutron measurement can be concluded to about 0.2 μg-U²³⁵/cps. Thus, it is confirmed that the delayed neutron detection is sensitive enough to monitor the fuel failure, and that the neutron count rate is high enough for stable signal with short counting time.     

Indirect measurement of inelastic cross section of relativistic protons in Pb target

The inelastic cross section of relativistic protons in Pb was determined indirectly by measuring the neutron distribution along a Lead spallation neutron source. The spallation neutron source was irradiated by 1, 1.5 and 2 GeV protons. The experimental results were obtained using passive methods. By the use of the beam attenuation coefficient, deduced by a fitting procedure of experimental data, the inelastic cross section of protons in Pb was determined.     

Theoretical cross sections of Bi,Th, U and U on deuteron-induced reactions

There are many application fields for fast neutrons. The main application fields of the fast neutrons are accelerator-driven sub-critical systems (ADS) and fusion–fission (hybrid) reactor systems for fission energy production. Thorium (Th) and uranium (U) are nuclear fuels in fusion–fission (hybrid) reactor systems and bismuth (Bi) is also the target nucleus in the ADS reactor systems. In this study, neutron production cross sections produced by (d, xn) reactions for spallation targets such as ²⁰⁹Bi, ²³²Th, ²³⁵U and ²³⁸U have been investigated. New evaluated hybrid model and geometry dependent hybrid model have been used to calculate the pre-equilibrium neutron production cross sections. For the reaction equilibrium component, Weisskopf–Ewing model calculations have been preferred. The obtained results have been discussed and compared with the available experimental data and found in agreement with each other.     

Evaluation of the fraction of delayed photoneutrons for TMSR-SF1

The 10 MW_(th) solid-fueled thorium molten salt reactor(TMSR-SF1) is a FLi Be salt-cooled pebble bed reactor to be deployed in 5–10 years, designed by the TMSR group. Due to a large amount of beryllium in the core, the photoneutrons are produced via(γ , n) reactions.Some of them are generated a long time after the fission event and therefore are considered as delayed neutrons. In this paper, we redefine the effective delayed neutrons into two fractions: the delayed fission neutron fraction and the delayed photoneutron fraction. With some reasonable assumptions, the inner product method and the k-ratio method are adopted for studying the effective delayed photoneutron fraction. In the k-ratio method, the Monte Carlo code MCNP6 is used to evaluate the effective photoneutron fraction as the ratio between the multiplication factors with and without contribution of the delayed neutrons and photoneutrons. In the inner product method, with the Monte Carlo and deterministic codes together, we use the adjoint neutron flux as a weighting function for the neutrons and photoneutrons generated in the core. Results of the two methods agree well with each other, but the k-ratio method requires much more computing time for the same precision.