Reactor Neutron Spectrum Research Articles

Independent isomeric yield ratios of fission products in the epi-cadmium neutron induced fission of 235U

In the epi-cadmium neutron induced fission of 235U, independent isomeric yield ratios (IR) of fission products 130,132Sb, 131,133Te, 134,136I, 135Xe and 138Cs have been measured by using an off-line gamma-ray spectrometric technique. The average neutron energy of the epi-cadmium reactor neutron spectrum is 1.9 MeV. From the IR values, the root mean square fragment angular momenta (JRMS) were deduced by using spin dependent statistical model analysis. The IR and JRMS values of considered fission products in the epi-cadmium neutron induced fission of 235U were compared with the literature data in the thermal neutron induced fission of 235U to examine the influence of excitation energy on nuclear structure effect.

Computational Optimization of 133m Xe Production in the Washington State University TRIGA Reactor

Here we present a new method of irradiating 132Xe capsules with neutrons to produce 133mXe gas standards that are used for radiation detector calibration at radioxenon measurement laboratories in support of the Comprehensive Test Ban Treaty (CTBT). This method is designed to maximize the production of 133mXe compared to 133Xe, both of which are competing products from the 132Xe(n, g) reaction. The 133mXe is produced at a much higher fraction for high-energy neutron absorptions in 132Xe (~50% for fast neutrons versus ~11% for thermal neutrons). We performed “spectral tuning” of the Washington State University (WSU) TRIGA reactor neutron spectrum inside the 132Xe ampules to maximize the number of fast neutrons and minimize the number of thermal neutrons available for 132Xe absorption. Spectral tuning analysis, done with Monte Carlo simulations, provided valuable insights into a future final design for a 132Xe irradiation capsule. With no spectral tuning, the fractional yield of 133mXe in the WSU reactor was ~11.7%. By surrounding the 132Xe capsule with a 0.5-cm-thick layer of tungsten and a 2.83-cm layer of europium (III) oxide and placing it in the reactor’s cadmium rotator tube next to the fuel elements, the fractional yield of 133mXe can be increased to 24.6%, a 111% increase in yield. Thus, by improving the fractional yield of 133mXe through spectral tuning, the CTBT will have better quality gas standards to use for radioxenon detector calibration to assist in the CTBT’s mission.

Independent isomeric yield ratios of fission products in the epi-cadmium neutron induced fission of 239Pu

Measurement of independent isomeric yield ratios (IR) of 128,130,132Sb, 131,133Te, 132.134,136I, 135Xe and 138Cs have been carried out for the first time in the epi-cadmium neutron induced fission of 239Pu by using an off-line gamma-ray spectrometric technique. The average neutron energy (<En>) of the epi-cadmium reactor neutron spectrum is 1.9 MeV. From the IR values, root mean square fragment angular momenta (JRMS) were deduced by using spin dependent statistical model analysis. Effect of nuclear structure on JRMS values was examined. The present data in the epi-cadmium neutron induced fission of 239Pu were compared with the similar data in the thermal neutron induced fission of 239Pu to examine the role of excitation energy on JRMS values.

Independent isomeric yield ratios of fission products in the epi-cadmium neutron induced fission of 233U

The independent isomeric yield ratios (IR) of 128,130,132Sb, 131,133Te, 132,134,136I, 135Xe and 138Cs have been measured in the epi-cadmium neutron induced fission of 233U by using an off-line gamma-ray spectrometric technique. The average neutron energy of the epi-cadmium reactor neutron spectrum is 1.9 MeV. The root mean square fragment angular momenta (JRMS) were deduced from the IR values by using spin dependent statistical model analysis. The IR and JRMS values of considered fission products in the epi-cadmium neutron induced fission of 233U were compared with the literature data in the thermal neutron induced fission of 233U to examine the influence of excitation energy on nuclear structure effect.

Independent isomeric yield ratios of fission products in the epi-cadmium neutron induced fission of 238Pu

In the epi-cadmium neutron induced fission of 238Pu, independent isomeric yield ratios (IR) of 128,130,132Sb, 131,133Te, 134,136I, 135Xe and 138Cs have been measured by using an off-line γ-ray spectrometric technique. The average neutron energy of the epi-cadmium reactor neutron spectrum is 1.9 MeV. From the IR values, root mean square fragment angular momenta (JRMS) were deduced by using spin dependent statistical model analysis. The JRMS values from the present work in the epi-cadmium neutron induced fission of 238Pu were compared with the literature data in the thermal neutron induced fission of 238Pu to examine the role of excitation energy. Effect of nuclear structure on JRMS values was examined.

Independent isomeric yield ratios of fission products in the epi-cadmium neutron induced fission of 241Pu

Abstract Independent isomeric yield ratios (IR) of 130,132Sb, 131,133Te, 134,136I, 135Xe and 138Cs have been measured in the epi-cadmium neutron induced fission of 241Pu by using an off-line gamma-ray spectrometric technique. The average neutron energy (&lt;E n&gt;) of the epi-cadmium reactor neutron spectrum is 1.9 MeV. From the IR values, root mean square fragment angular momenta (J RMS) were deduced by using spin dependent statistical model analysis. Effect of nuclear structure on J RMS values was examined. The present data in the epi-cadmium neutron induced fission of 241Pu were compared with the similar data in the thermal neutron induced fission of 241Pu to examine the role of excitation energy on fragment angular momentum.

Measurement of spectrum-averaged cross sections of the (n,p) and (n,n′) reactions on strontium by fast neutrons of a TRIGA reactor: comparison with integrated data from excitation functions of various data libraries

Abstract Integral cross sections of the reactions 84Sr(n,p)84Rb, 86Sr(n,p)86Rb and 87Sr(n,n′)87mSr, over the fast neutron spectrum of a TRIGA reactor above 0.5 MeV, were measured for the first time. The results were used for integral tests of the evaluated excitation functions in the data libraries, ENDF/B-VIII.0, ROSFOND-2010, JEFF-3.3, JENDL-5 and TENDL-2021. The integral measurements are generally consistent with the integrated values within 12 %, except for a few cases. The two data libraries ENDF/B-VIII.0 and ROSFOND-2010 appear to need some improvement with regard to the reactions investigated in this work.

Status of the engineering design of the IFMIF-DONES high energy beam transport line and beam dump system

IFMIF-DONES plant [1] (International Fusion Materials Irradiation Facility – DEMO Oriented Neutron Source) is currently being developed under the framework of a work package of the EUROfusion Consortium. It will be a facility located at the south of Spain in Granada. Its objective and main goal is the testing and qualification of fusion materials by the generation of a neutron flux with a broad energy distribution covering the typical neutron spectrum of a (D-T) fusion reactor. This is achieved by the Li (d,xn) nuclear reactions occurring in a liquid lithium target where a 40 MeV at 125 mA deuteron beam with a variable beam footprint between 200 mm × 50 mm and 100 mm × 50mm collides. The Accelerator Systems is in charge of providing such high energy deuterons to produce the required neutron flux. The High Energy Beam Transport line (HEBT) is the last subsystem of the IFMIF-DONES accelerator, and its main functions are to guide the deuteron beam towards the liquid lithium target and to shape it with the required rectangular reference beam footprints. The HEBT system includes a Beam Dump devoted to stop the beam during commissioning and start-up phases. The present work details the present status of the HEBT engineering design, including beam dynamics, vacuum configuration, radioprotection, beam diagnostics devices and remote handling analyses performed detailing the layout and integration of required components through-out the beamline.

Impact of reactor neutron spectrum on measured spectrum averaged cross sections

The cross section averaged over 235U thermal-neutron induced fission spectrum is a fundamental quantity that can be used in evaluation and validation of nuclear data. Many experiments focused on the determination of Spectrum Averaged Cross Sections (SACS) in 235U(nth,f) Prompt Fission Neutron Spectrum (PFNS) in light water reactors using enriched uranium fuel. In these reactors, there is already some amount of water moderator between the uranium fuel and the irradiated sample. Due to the decrease of hydrogen cross-section with neutron energy, the high energy tail of the reactor spectrum in cores with water moderator may be harder than the pure PFNS. This paper aims to compare the shape of the actual reactor spectrum in various core positions of a research light-water reactor differing each from other by the effective water thickness. The spectrum shape is determined both by calculations and experimentally using various high energy threshold reactions.The impact of the photo-nuclear reactions (γ,n) competing with (n,2n) in production of the same residual nucleus was shown to be less than a percent for most of studied dosimeters. An important exception was found for 197Au(n,2n)196Au dosimeter irradiated in the outcore channel where a notable photo-neutron contribution to the production of 196Au is caused by the neutron production from the high energy γ-rays from thermal-neutron capture in 54Fe. The corresponding ENDF/B-VIII.0 data turned out to underestimate such γ-yield by 40 % in comparison with ENDF/B-VI.8. This has improved but however not resolved the disagreement between our measurement and calculations. The remaining deficiency was attributed to the underestimation of the evaluated cross section IAEA/PD-2019 for the 197Au(γ,n) cross section near the reaction threshold. The later was confirmed by comparison with existing measured data.

DONES Systems identification and requirements allocation

• Systems identification of IFMIF-DONES for work distribution and responsbility allocation. • System-of-Systems to Group-of-Systems type mapping. • Alignment of Systems identification and requirement allocation. The mission of the DONES program is to develop a database of fusion-like neutron irradiation effects in the materials required for the construction of the Demonstration Fusion Power Reactor (DEMO) for benchmarking of radiation response of materials. In order to do this, the DONES facility will be built to provide a neutron source producing high-energy neutrons at sufficient intensity and irradiation volume. DONES will generate a neutron flux with a broad energy distribution covering the typical neutron spectrum of a (d-t) fusion reactor. This is achieved by utilizing Li(d,xn) nuclear reactions taking place in a liquid Li target when bombarded by a deuteron beam. The energy of the deuterons (40 MeV) and the current of the accelerator (125 mA) have been tuned to maximize the neutron flux to get irradiation conditions comparable to those in the first wall of a fusion power reactor. For the success of this mission, the needs of all the stakeholders shall be very clearly stated, considering the whole lifecycle of the facility. Once the needs are clear, they shall be translated into specific requirements, following a top-down approach, to be allocated to the different systems or groups-of-systems (GoS). To do this, it is essential to properly identify which parts shall be treated as a System or GoS, as they shall be designed accordingly, and the requirements allocated in consequence.

A Multi-Scale Simulation Study of Irradiation Swelling of Silicon Carbide.

Silicon carbide (SiC) is a promising structural and cladding material for accident tolerant fuel cladding of nuclear reactor due to its excellent properties. However, when exposed to severe environments (e.g., during neutron irradiation), lattice defects are created in amounts significantly greater than normal concentrations. Then, a series of radiation damage behaviors (e.g., radiation swelling) appear. Accurate understanding of radiation damage of nuclear materials is the key to the design of new fuel cladding materials. Multi-scale computational simulations are often required to understand the physical mechanism of radiation damage. In this work, the effect of neutron irradiation on the volume swelling of cubic-SiC film with 0.3 mm was studied by using the combination of molecular dynamics (MD) and rate theory (RT). It was found that for C-vacancy (CV), C-interstitial (CI), Si-vacancy (SiV), Si-interstitial (SiI), and Si-antisite (SiC), the volume of supercell increases linearly with the increase of concentration of these defects, while the volume of supercell decreases linearly with the increase of defect concentration for C-antisite (CSi). Furthermore, according to the neutron spectrum of a certain reactor, one RT model was constructed to simulate the evolution of point defect under neutron irradiation. Then, the relationship between the volume swelling and the dose of neutrons can be obtained through the results of MD and RT. It was found that swelling typically increases logarithmically with radiation dose and saturates at relatively low doses, and that the critical dose for abrupt transition of volume is consistent with the available experimental data, which indicates that the rate theory model can effectively describe the radiation damage evolution process of SiC. This work not only presents a systematic study on the relationship between various point defect and excess volume, but also gives a good example of multi-scale modelling through coupling the results of binary collision, MD and RT methods, etc., regardless of the multi-scale modelling only focus on the evolution of primary point defects.

Computational and experimental optimization of 135Xe production in calibration sources

Here we present a new method of irradiating 134Xe capsules to produce 135Xe gas standards which maximize the ratio of 135Xe to 133Xe production due to (n,g) and (n,2n) reactions, respectively. We performed “Spectral tuning” of the University of Utah TRIGA Reactor (UUTR) neutron spectrum to increase the length of time that 135Xe dominates undesirable 135Xe in the sample, so that the capsules – used for calibration and quality control testing of Xe gas detection equipment in support of the Comprehensive Test Ban Treaty (CTBT) – will remain viable for longer periods post-irradiation. Moreover, optimized methods of computation and analysis were developed yielding improved computational efficiency over standard Monte Carlo approaches. These methods provided valuable insight into the final design and manufacture of new, ex-core Teflon irradiation chambers tested in the UUTR. The methods of computation and analysis, as well as the physical irradiation chamber designs, were derived such that they could be readily applied to any reactor for spectral tuning of a specific reactor's flux profile. Results of the physical experiments employing the optimized irradiation chamber designs demonstrated sample viability time improvements of over 60% when compared to conventional, un-optimized methods of gas sample generation. Thus, use of these methods enhance both the CTBT-related calibrations and performance testing, and the continued stability of the CTBT monitoring network overall.

RBSM-based model for prediction of radiation-induced volumetric expansion of concrete aggregates

A new comprehensive 2D RBSM-based numerical model for prediction of radiation-induced volumetric expansion (RIVE) of aggregates depending on rock mineral composition is explained in detail in this study. The new model combines the existing model of RIVE of minerals with the new rules, which take into account the anisotropic behavior of minerals. The numerical model also takes into account irradiation temperature, neutron fluence and reactor neutron spectrum. The overall coefficient of determination for the numerical model validation of 0.960 and the residual standard error of 0.961% make the described numerical model suitable for reliable prediction of RIVE of aggregates.

NEUTRON FIELD MOCK-UP DEVELOPMENT FOR THE FLUORIDE SALT REACTORS NEUTRONIC RESEARCH

Experimental work at the LR-0 reactor was carried out to determine the possibility of the mock-up neutron field creation for the fluoride salt-based reactors. Previous calculations and experiments have shown that the fast part of the molten salt reactor (MSR) spectrum is shaped by fluorine and even the Teflon material is suitable for neutronics in that energy range. Properly selected spectrum indices can describe the neutron field of the MSR in fast thermal and intermediate parts of the spectrum. Current research has focused on a deeper study of the possibility of using the filtered thermal and intermediate neutron spectrum of the experimental light water reactor for the physics of fluoride salt-based reactor. LiF-BeF2 (FLIBE) capsule and teflon cylinders are used as a spectrum filters in the LR-0 reactor. Measured results show acceptable C/E-1 agreement in the reaction rates and satisfactory agreement for usage of the FLIBE filtered neutron spectrum determined by the 181Ta(n, γ) monitor as a mock-up in thermal to intermediate energy in the fluoride high-temperature reactor (FHR). Concerning MSR, intermediate spectrum can be reproduced to some extent but not as good as in the case of FHR.

EFFECT OF RADIATION HEATING OF CONSTRUCTION MATERIALS OF AN EXPERIMENTAL DEVICE WITH A NEUTRON CONVERTER ON THEIR TEMPERATURE REGIME

The paper considers an experimental device with a neutron converter [1], which provides a change in the neutron spectrum of the IGR reactor from thermal to fast. Successful testing requires a number of thermophysical calculations that sufficiently accurately simulate the thermal state of the experimental device and make it possible to justify the safe conditions of the experiment. The paper presents the results of non-stationary calculations of the thermal state of an experimental device with a neutron converter with energy release both in fuel and in structural materials that simulate nuclear fission energy in fuel elements and the effect of instantaneous and delayed gamma radiation on structural materials.

Metallic impurities induced corrosion of a Ni-26W-6Cr alloy in molten fluoride salts at 850 oC

The effect of metallic impurities on corrosion of a Ni-26W-6Cr alloy in molten FLiNaK salts at 850 °C was investigated. Results show that the Ni-26W-6Cr alloy exhibits the same excellent compatibility with fluoride salts as Hastelloy N alloy. Fe impurities in molten salts accelerate the diffusion of Cr in the alloy, which determines the corrosion degree of the alloy. Unexpected depletion of W from the alloy was observed, which mostly results from the presence of Mo impurity in the salts. Mo impurity should be limited because of the significant effect of W on the neutron spectra of reactors.

Separation of Uranium and Zirconium in Alkali Chloride Melts Using Liquid Metal Cathodes

Alkali chloride melts are considered as prospective working media for non-aqueous pyrochemical reprocessing of spent nuclear fuels (SNFs). Separation of fissile materials from fission products in pyrochemical reprocessing can be achieved electrochemically and amongst all the fission products zirconium has the closest electrochemical properties to uranium. Uranium and plutonium fission produces several zirconium isotopes (from Zr-91 to Zr-97) and depending on the reactor neutron spectrum, nuclear fuel type, burnup and cooling time SNF arriving for reprocessing can contain ca. 5–13 kg of zirconium (as the fission product excluding cladding) per ton. Electrochemical separation of uranium and zirconium in fused salts is a challenging task.The present work was devoted to studying the electrochemical behavior of zirconium and uranium in 3LiCl–2KCl based melts using cyclic voltammetry and cathodic polarization. The experiments were performed in LiCl–KCl–ZrCl4, LiCl–KCl–UCl4, LiCl–KCl–UCl3 and LiCl–KCl–ZrCl4–UCl4 melts.On a solid tungsten electrode zirconium(IV) ions were reduced to Zr(0) in two stages and the metal deposition potentials were between –2.17 and –2.07 V (at 532–637 oC) vs. Cl–/Cl2 couple. Changing the solid electrode to liquid zinc, gallium or gallium–zinc eutectic alloy (3.64 wt. % Zn) resulted in significant shift of zirconium deposition potential in the positive direction. Examples of the polarization curves are shown in the Fig. Polarization measurements performed in LiCl–KCl–ZrCl4–UCl4 melt on the Ga–Zn electrode showed that reduction of U(IV) to U(III) and deposition of zirconium occurred at very similar potentials and deposition potential of uranium was significantly more negative. Separation factor for uranium/zirconium couple was also determined.Fig. Polarization of Ga–Zn eutectic alloy cathode in LiCl–KCl–UCl4 (524 oC) and LiCl–KCl–ZrCl4 (550 oC) melts. Figure 1

Possibility of Burning Americium in Fast Reactors

The possibilities of burning americium from spent nuclear fuel of thermal reactors in the operating BN-800 are examined. It is shown that because of the small fission cross-section in the neutron spectrum of a fast reactor all of the americium cannot be directly converted into fission products. This can be done only by transmutation of americium into plutonium, which must be separated and directed into fuel, where it will burn. In a reactor only 15–20% of the americium is transmuted, and the remaining americium must be returned into the reactor for transmutation. The volume of the recycled americium can be significantly reduced by burning in facilities possessing a moderator (zirconium hydride) and the need for burning it can be eliminated by using plutonium with a high content of the isotope 241Pu as the fuel of fast reactors, where this isotope fissions efficiently.

Integral measurement of spectrum-averaged cross sections of a few threshold reactions induced by fast neutrons of a TRIGA reactor: comparison with integrated data from excitation functions given in various data libraries

Abstract Integral cross sections of the reactions 24Mg(n,p)24Na, 27Al(n,p)27Mg, 27Al(n,α)24Na, 58Ni(n,d + np)57Co and 60Ni(n,p)60Co were measured for the first time using the fast neutron spectrum of a TRIGA reactor extending from 0.5 to 20 MeV. The values obtained in this work were comparable with the recommended values for a pure 235U prompt fission spectrum. The measured integral value was utilized for integral test of excitation function of each reaction given in some data libraries, namely ENDF/B-VIII.0, TENDL-2017, IRDFF-1.05 and ROSFOND-2010. The integral measurements are generally consistent with the integrated values within 5 %, except for a few cases, e. g. the reaction 60Ni(n,p)60Co, where the data libraries appear to need improvement.

Study of transmutation, gas production, and displacement damage in chromium for fusion neutron spectrum

In the present work, gas production, transmutation and displacement damage in chromium are studied for fusion neutron spectra. The nuclear cross-section data are calculated with the TALYS-1.8 code. Transmutated isotopes in chromium, including the radioactive ones, are identified for typical fusion reactor neutron spectrum. Gas production per atom (He and H production) are also predicted for typical fusion neutron spectrum. Molecular dynamics (MD) simulations of damage cascade caused by the self-recoil in chromium are carried out for up to 200 keV damage energies using the LAMMPS code. Time evolution of interstitials and vacancies are studied and discussed. Based on the performed MD simulations, constant parameters of the athermal-recombination corrected dpa (arc-dpa) method are calibrated. Displacement damage cross-section of chromium are calculated using the Norgett-Robinson and Torrens (NRT) and arc-dpa approaches and later used to predict the values of displacement per atoms (dpa) in chromium for typical fusion reactor neutron environment.