Determination of the 238U capture to total fission ratio in alternate depleted uranium/polyethylene shells with D-T neutrons

Aiming at checking the conceptual design of the subcritical blanket in the fusion–fission hybrid reactor, an integral experiment was carried out on an alternate depleted uranium/polyethylene-shell setup with D-T neutrons using activation technique. 18 depleted uranium foils were placed at 90° direction to the incident D beam, and the distribution of the ²³⁸U capture to total fission ratio was determined by measuring the 277.6 keV γ ray generated by neutron capture of ²³⁸U and the 293.3 keV γ ray generated by fission of ²³⁵U and ²³⁸U. The ratios were generally between 1 and 2 in the depleted uranium shells, with relative uncertainties between 3.0% and 5.5%. The ratios were calculated by the MCNP4B code employing ENDF/B-VI nuclear data library, the discrepancies between calculations and experiments were all within 6%, and the average calculation to experiment(C/E) ratio was 0.998.     

Parameters optimization in a hybrid system with a gas dynamic trap based neutron source

A task of long-lived transuranic isotopes utilization is considered to be one of the urgent problems for the nuclear reactor technology. Using sub-critical hybrid systems is a possible solution of the problem. Budker Institute of Nuclear Physics SB RAS together with Nuclear Safety Institute RAS is working on a hybrid system with a neutron source based on the gas dynamic trap and sub-critical fuel blanket. This article presents the results obtained from a series of numerical experiments aimed at estimating the optimal system. Particularly, maximum neutron source emission rate has been estimated to reach 1 × 10¹⁸–2 × 10¹⁸ neutrons/s at the input parameters typical for such a system. Pb–Bi buffer zone impact on integral characteristics of fuel blanket has been considered. Decrease in amount of secondary fission neutrons as the result of buffer zone thickening has been revealed.The codes developed to conduct the investigations are also described in the article. The first one, GENESYS, is a zero-dimensional code aimed at modelling plasma processes in the gas dynamic trap. The second one, NMC (Neutron Monte-Carlo), is a Monte-Carlo particle transport code and is developed as a multipurpose tool for neutron transport calculation.     

Measurements and model calculations of activation cross sections for 232Th(n,2n)231Th reaction between 13.57 and 14.83 MeV neutrons

In this study, activation cross sections were measured for the reaction of ²³²Th(n,2n)²³¹Th (T_1/2 = 25.5 h) by using neutron activation technique at six different neutron energies from 13.57 and 14.83 MeV. Neutrons were produced via the ³H(²H,n)⁴He reaction using SAMES T-400 neutron generator. Irradiated and activated high purity Thorium foils were measured by a high-resolution γ-ray spectrometer with a high-purity Germanium (HpGe) detector. In cross section measurements, the corrections were made for the effects of γ-ray self-absorption in the foils, dead-time, coincidence summing, fluctuation of neutron flux, low energy neutrons. For this reaction, statistical model calculation, which the pre-equilibrium emission effects were taken into consideration, were also performed between 13.57 and 14.83 MeV energy range. The cross sections were compared with previous works in literature, with model calculation results, and with evaluation data bases (ENDF/B-VII, ENDF/B-VI, JEFF-3.1, JENDL-4.0, JENDL-3.3, and ROSFOND-2010).     

Feasibility of a multi-purpose demonstration neutron source based on a compact superconducting spherical tokamak

Tokamak neutron sources would allow near term applications of fusion such as fusion–fission hybrid reactors, elimination of nuclear wastes, production of radio-isotopes for nuclear medicine, material testing and tritium production. The generation of neutrons with fusion plasmas does not require energetic efficiency; thus, nowadays tokamak technologies would be sufficient for such purposes. This paper presents some key technical details of a compact (～1.8 m³ of plasma) superconducting spherical tokamak neutron source (STNS), which aims to demonstrate the capabilities of such a device for the different possible applications already mentioned. The T-11 transport model was implemented in ASTRA for 1.5 D simulations of heat and particle transport in the STNS core plasma. According to the model predictions, total neutron production rates of the order of ～10¹⁵ s^?1 and ～10¹³ s^?1 can be achieved with deuterium/tritium and deuterium/deuterium respectively, with 9 MW of heating power, 1.4 T of toroidal magnetic field and 1.5 MA of plasma current. Engineering estimates indicate that such scenario could be maintained during ～20 s and repeated every ～5 min. The viability of most of tokamak neutron source applications could be demonstrated with a few of these cycles and around ～100 cycles would be required in the worst cases.     

Experimental prompt fission neutron “sawtooth” data described by the “point by point” model

The experimental data concerning the prompt fission neutron number as function of the fission fragment mass (currently named “sawtooth” data) when they exist, allow a more refined verification of the “point by point” model and in the same time the validation of the methods used to determine the model parameters corresponding to the fission fragment pairs covering the entire fission fragment range.The prompt neutron multiplicities of each fission fragment pair provided by the “point by point” model are compared with the experimental data concerning the sum of neutrons emitted by the light and heavy fragments forming a pair obtained from experimental sawtooth data.The available fission fragment pair multiplicity and sawtooth experimental data for the ^233,235,238U(n, f) and ²³⁷Np(n, f) reactions as well as for the spontaneous fission of ²⁵²Cf are well described by the “point by point” model results, proving again that this treatment is a powerful tool for prompt fission neutron multiplicity and spectrum evaluation purposes.     

Nondestructive Magnetic Adaptive Testing of nuclear reactor pressure vessel steel degradation

Inspection of neutron-irradiation-generated degradation of nuclear reactor pressure vessel steel (RPVS) is a very important task. In ferromagnetic materials, such as RPVS, the structural degradation is connected with a change of their magnetic properties. In this work, applicability of a novel magnetic nondestructive method (Magnetic Adaptive Testing, MAT), based on systematic measurement and evaluation of minor magnetic hysteresis loops, is shown for inspection of neutron irradiation embrittlement in RPVS. Three series of samples, made of JRQ, 15CH2MFA and 10ChMFT type steels were measured by MAT. The samples were irradiated by E > 1 MeV energy neutrons with total neutron fluence of 1.58 × 10¹⁹–11.9 × 10¹⁹ n/cm². Regular correlation was found between the optimally chosen MAT degradation functions and the neutron fluence in all three types of the materials. Shift of the ductile–brittle transition temperature, ΔDBTT, independently determined as a function of the neutron fluence for the 15CH2MFA material, was also evaluated. A sensitive, linear correlation was found between the ΔDBTT and values of the relevant MAT degradation function. Based on these results, MAT is shown to be a promising (at least) complimentary tool of the destructive tests within the surveillance programs, which are presently used for inspection of neutron-irradiation-generated embrittlement of RPVS.     

Overview of the US–Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. A recent MIT report (PSFC/RR-10-4, An assessment of the current data affecting tritium retention and its use to project towards T retention in ITER, Lipschultz et al., 2010) summarizes the observations from high-energy ion bombardment studies and illustrates the saturation trend in deuterium concentration due to damage from ion irradiation in tungsten and molybdenum above 1 displacement per atom (dpa). While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in plasma facing components (PFCs), it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high primary knock-on atom (PKA) energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influences the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron-irradiation. Under the framework of the US–Japan TITAN program, tungsten samples (99.99 at.% purity from A.L.M.T. Co.) were irradiated by fission neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory (ORNL), at 50 and 300 °C to 0.025, 0.3, and 2.4 dpa, and the investigation of deuterium retention in neutron-irradiated tungsten was performed in the Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe²⁺, 20 MeV W²⁺, and 700 keV H^?) with that from neutron-irradiated tungsten to identify the similarities and differences among them.     

Measurement of thermal neutron cross section and resonance integral for 165Ho(n, γ)166gHo reaction by the activation method

The thermal neutron cross section and the resonance integral of the reaction ¹⁶⁵Ho(n, γ)^166gHo were measured by the activation method using ⁵⁵Mn(n,γ)⁵⁶Mn monitor reaction. The sufficiently diluted MnO₂ and Ho₂O₃ samples with and without a cylindrical Cd case were irradiated in an isotropic neutron field of the ²⁴¹Am–Be neutron sources. The γ-ray spectra from the irradiated samples were measured with a calibrated n-type high purity Ge detector. Thus, the thermal neutron cross section for ¹⁶⁵Ho(n,γ)^166gHo reaction has been determined to be 59.2 ± 2.5 b relative to the reference thermal neutron cross section value of 13.3 ± 0.1 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction, and it generally agrees with the recent measurements within about 1 to 12%. The resonance integral has also been measured relative to the reference value of 14.0 ± 0.3 b for the ⁵⁵Mn(n,γ)⁵⁶Mn reaction using an epithermal neutron spectrum of the ²⁴¹Am–Be neutron source. The resonance integral for ¹⁶⁵Ho(n, γ)^166gHo reaction obtained was 667 ± 46 b at a cut-off energy of 0.55 eV for 1 mm Cd thickness. The existing experimental and evaluated data for the resonance integral are distributed from 618 to 752 b. The present resonance integral value agrees with most of the previously reported values obtained by ¹⁹⁷Au standard monitor within the limits of error.     

Investigation of efficient 131I production from natural uranium at Tehran research reactor

Iodine-131, which has a half-life of 8.05 days, is the one of the most widely used radionuclides in medical diagnosis and treats some diseases of thyroid gland. Optimization of ¹³¹I production in Tehran research reactor (TRR) was studied by two different methods. Primarily, standard nuclear codes such as ORIGEN, WIMS and CITATION were applied and then analytical solutions technique was followed.Calculated results and experimental works in the bench scale indicate that, by irradiation of 100 g natural Uranium (UO₂) for 100 h at 3.5 × 10¹³ (n’s/cm² s) thermal neutron flux in the TRR, one can produce about 5 Ci of ¹³¹I for medical purposes, on the other hand can produce very useful radionuclides like ⁹⁹Mo and ¹³³Xe in one batch irradiation in the unique production line.     

The thermal neutron diffusion cooling coefficient in polyethylene

The diffusion cooling coefficient C for thermal neutrons in polyethylene at 20°C has been determined theoretically. Granada's Synthetic Model of the scattering law has been applied to describe the interaction of neutrons with polyethylene. Two approximations of the neutron energy distribution in finite homogeneous systems have been used. The result of the calculation using a rough approximation is C_B=2160 cm⁴ s⁻¹. According to a more advanced formalism which follows Nelkin's analysis of the neutron pulse decay in a finite medium, applying the diffusion theory with transport correction, the value obtained is C=2916 cm⁴ s⁻¹.     

Design of a large irradiation channel at MNSR facility in Ghana

In the design of new slant tube for large sample irradiation in the Ghana Research Reactor-1 facility, Monte Carlo N-Particle Code version 5 (MCNP-5) was employed to simulate the neutron flux profile of the new design. The results show that the neutron flux peaks at different points, at an average thermal neutron flux of (1.1406 ± 0.0046) × 10¹¹, (1.1849 ± 0.0047) × 10¹¹ and (1.0580 ± 0.0044) × 10¹¹ n cm^?2 s^?1 around the reactor vessel. The first two peaks happened to coincide with pneumatic transfer pipes in the pool, but the third peak happened to be in line with the slant tube position. It was observed that as the diameter of the tube varies from 3.90 cm to 23.40 cm, the average thermal neutron flux decreased exponentially from (1.1849 ± 0.0047)10¹¹ n cm^?2 s^?1 to (3.3241 ± 0.0100) × 10¹⁰ n cm^?2 s^?1. The average thermal neutron flux decreases exponentially along the diameter of the designed slant tube from (1.0366 ± 0.0042) × 10¹¹ n cm^?2 s^?1 to (9.7396 ± 0.0136) × 10⁹ n cm^?2 s^?1. From the results, it is evident that a slant tube of diameter 15.00 cm can be installed at the original slant tube position for large sample irradiation.     

Activation analysis of coolant water in ITER blanket and divertor

Coolant water in blankets and divertor cassettes will be activated by neutrons during ITER operation. ¹⁶N and ¹⁷N are determined to be the most important activation products in the coolant water in terms of their impact on ITER design and performance. In this study, the geometry of cooling channels in blanket module 4 was described precisely in the ITER neutronics model ‘Alite-4’ based on the latest CAD model converted using MCAM developed by FDS Team. The ¹⁶N and ¹⁷N concentration distribution in the blanket, divertor cassette and their primary heat transport systems were calculated by MCNP with data library FENDL2.1. The activation of cooling pipes induced ¹⁷N decay neutrons was analyzed and compared with that induced by fusion neutrons, using FISPACT-2007 with data library EAF-2007. The outlet concentration of blanket and divertor cooling systems was 1.37 × 10¹⁰ nuclide/cm³ and 1.05 × 10¹⁰ nuclide/cm³ of ¹⁶N, 8.93 × 10⁶ nuclide/cm³ and 0.33 × 10⁵ nuclide/cm³ of ¹⁷N. The decay gamma-rays from ¹⁶N in activated water could be a problem for cryogenic equipments inside the cryostat. Near the cryostat, the activation of pipes from ¹⁷N decay neutrons was much lower than that from fusion neutrons.     

Evaluation of gamma-ray and neutron energy for area monitoring system in the IFMIF/EVEDA accelerator building

The engineering validation of the IFMIF/EVEDA prototype accelerator, up to 9 MeV by supplying the deuteron beam of 125 mA, will be performed at the BA site in Rokkasho. A design of this area monitoring system, comprising of Si semiconductors and ionization chambers for covering wide energy spectrum of gamma-rays and ³He counters for neutrons, is now in progress. To establish an applicability of this monitoring system, photon and neutron energies have to be suppressed to the detector ranges of 1.5 MeV and 15 MeV, respectively. For this purpose, the reduction of neutron and photon energies throughout shield of water in a beam dump and concrete layer is evaluated by PHITS code, using the experimental data of neutron source spectra. In this article, a similar model using the beam dump structure and the position with a degree of leaning for concrete wall in the accelerator vault is used, and their energy reduction including the air is evaluated. It is found that the neutron and photon flux are decreased by 10⁴-order by employing the local shields using concrete and polyethylene around beam dump, and the photon energy can be suppressed in the low energy.     

Measurement of (n, n′) reaction cross-sections of 79Br, 90Zr, 197Au and 207Pb with pulsed d-D neutrons

Activation cross-sections for the (n, n′) reaction were measured by means of the activation method at neutron energies of 3.1 and 2.54 MeV using a pulsed neutron beam. The target nuclei were ⁷⁹Br, ⁹⁰Zr, ¹⁹⁷Au, and ²⁰⁷Pb whose half-lives were between 0.8 and 8 s. The value of the ⁹⁰Zr(n, n′) ^90mZr reaction was obtained for the first time. In order to confirm the pulsed neutron beam measuring method, the cross-section data of ⁷⁹Br and ¹⁹⁷Au were compared with previous data obtained using a pneumatic sample transport system. The results of this comparison were in agreement within the range of experimental error. The d-D neutrons were generated by bombarding a deuterated titanium target with a 350-keV d⁺-beam at the 80° beam line of the Fusion Neutronics Source (FNS) at the Japan Atomic Energy Research Institute. In order to obtain reliable activation cross-sections, careful attention was paid to correct the efficiency for a volume source, and the self-absorption of gamma rays in irradiated samples. The systematics of the (n, n′) reaction at a neutron energy of 3.0 MeV, which can predict cross-section of (n, n′) reaction with an accuracy of 50%, was proposed for the first time on the basis of our data.     

Monte Carlo calculations of the nuclear effects of certain fluids in a hybrid reactor

This study analyzes the effects of certain heavy-metal-salt fluids on nuclear parameters in a fusion–fission hybrid reactor. Calculated parameters include the tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate, and fissile fuel breeding in the reactor's liquid first wall, blanket, and shield zones; gas production rates in the structural material of the reactor were calculated, as well. The fluid mixtures consisted of 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% UO₂, 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% NpO₂, and 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% UCO. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion–fission hybrid reactor system. A 3 cm wide beryllium (Be) zone was used for neutron multiplier between the liquid first wall and the blanket. The structural material used was 9Cr2WVTa ferritic steel, measuring 4 cm in width. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and the ENDF/B-VII.0 nuclear data library.     

Feasibility study of an intense D–T fusion source: “The New Sorgentina”

The international community agrees on the importance to build a large facility devoted to test and validate materials to be used in harsh neutron environments. Such a facility, proposed by ENEA, reconsiders a previous study known as “Sorgentina” but takes into account new technological development so far attained. The “New Sorgentina” Fusion Source (NSFS) project is based upon an intense D–T 14 MeV neutron source achievable with T and D ion beams impinging on 2 m radius rotating targets. NSFS produces about 1 × 10¹³ n cm⁻² s⁻¹ over about 50 cm³. Larger volumes of lower neutron flux will be available (e.g. for TBM experiments) as well as collimated channels to study some features of the ITER neutron camera. The NSFS facility will overcome problems related to the ion source and accelerating system, by means of an upgraded version of the JET–PINI ion beams. NSFS has to be intended as an European facility that may be realized in a few years, once provided a preliminary technological program devoted to the operation of the ion source in continuous mode, target heat loading/removal, target and tritium handling, inventory as well as site licensing. In this contribution, the main characteristics of NSFS project will be presented.     

Obtaining molybdenum-99 in the IRT-T research reactor using resonance neutrons

The yield of ⁹⁹Mo from the ⁹⁸Mo(n,γ)⁹⁹Mo reaction significantly depends of the energy spectrum of the neutron flux. It is well known that the cross-section for this reaction is about 130 mb, whereas the resonance integral of the reaction is 6.9 b. The aim of this work was to investigate the conditions that let to increase ⁹⁹Mo yield from the targets with natural and enriched isotope composition under irradiation by resonance neutrons at the IRT-T research reactor.The calculations of integrated cross-sections of all Mo isotopes in the region of the ⁹⁸Mo resonances showed that screening in the target with natural isotope composition by other isotopes is relatively small. So the ⁹⁸Mo in the natural mixture can be activated by resonance neutrons approximately in the same manner as pure ⁹⁸Mo.Experimental measurements of the ⁹⁸Mo(n,γ) effective cross-section using the MoO₃ sample with natural and enriched composition in the reactor channels with the beryllium moderator with the thickness of 20 up to 90 mm showed that the effective cross-sections in these channels reach the value of 700 mb. The contribution of the epithermal neutrons into the ⁹⁸Mo activity was 68% for the enriched targets and 78% for natural molybdenum, respectively.At that channel it is possible to produce ⁹⁹Mo with specific activity up to 3.4 Cu/g with samples of natural isotope composition and up to 15 Cu/g with enriched samples on the base of reactors with neutron flux of (1.7 × 10¹⁴ n/(cm² s)). Such ⁹⁹Mo specific activity is enough not only to realize extraction technologies production of ^99mTc, but to manufacture sorption generators of ^99mTc without wastes.     

Russian development of enhanced heat flux technologies for ITER first wall

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5 MW/m²) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr–SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V = 1–3 m/s, P = 2–3 MPa, T = 110–170 °C). The heat flux was gradually varied in the range of 1–10 MW/m² until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2 m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be–CuCrZr interface. Results of tile dimensions optimization (variable in the range 12 mm × 12 mm × 6 to 50 mm × 50 mm × 8 mm) obtained by the HHF (variable in the range of 3–8 MW/m²) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.     

Experimental study of onset of subcooled annular flow boiling

An experimental study on the onset of nucleate boiling (ONB) is performed for water annular flow to provide a systematic database for low pressure and velocity conditions. A parametric study has been conducted to investigate the effect of pressure, inlet subcooling, heat and mass flux on flow boiling. The test section includes a Pyrex tube with 21 mm inner diameter and a stainless steel (SS-304) rod with outer diameter of 6 mm. Pressure, heat and mass flux are in the range of 1.73 < P < 3.82 bar, 40 < q < 450 kW/m² and 70 < G < 620 kg/m² s, respectively. The results illustrate that inception heat flux is extremely dependent on pressure, inlet subcooling temperature and mass flux; for example in pressure, velocity and inlet subcooling as 3.27 bar, 230 kg/m² s and 41.3 °C; consequently q_w,ONB is 177.3 kW/m². In other case with higher inlet temperature of 71.5 °C and with P, 3.13 bar and G, 232 kg/m² s the inception heat flux reached to 101.6 kW/m². The data of ONB heat flux are over estimated from the existing correlation, and maximum deviation of wall superheat (ΔT_w,ONB) from correlations is 30%. Experimental data of inception heat flux are within 22% of that predicted from the correlation.     

2-D temperature distribution and heat flux of PFC in 2011 KSTAR campaign

KSTAR has reached a plasma current up to 630 kA, plasma duration up to 12 s, and has achieved high confinement mode (H-mode) in 2011 campaign. The heat flux of PFC tile was estimated from the temperature increase of PFC since 2010. The heat flux of PFC tiles increases significantly with higher plasma current and longer pulse duration. The time-averaged heat flux of shots in 2010 campaign (with 3 s pulse durations and I_p of 611 kA) is 0.01 MW/m² while that in 2011 campaign (with 12 s pulse duration and I_p of 630 kA) is about 0.02 MW/m². The heat flux at divertor is 1.4–2 times higher than that at inboard limiter or passive stabilizer. With the cryopump operation, the heat flux at the central divertor is higher than that without cryopump. The heat flux at divertor is proportional to, of course, the duration of H-mode. Furthermore, a software tool, which visualizes the 2D temperature distribution of PFC tile and estimates the heat flux in real time, is developed.