Gamma-ray Heating Research Articles

Gamma ray heating and neutrino cooling rates due to weak interaction processes on s d $sd$ -shell nuclei in stellar cores

Gamma ray heating and neutrino cooling rates, due to weak interaction processes, on \(sd\)-shell nuclei in stellar core are calculated using the proton neutron quasiparticle random phase approximation theory. The recent extensive experimental mass compilation of Wang et al. (Chin. Phys. C 36:1603, 2012), other improved model input parameters including nuclear quadrupole deformation (Raman et al. in At. Data Nucl. Data Tables 78(1):1–128, 2001; Moller et al. in At. Data Nucl. Data Tables 109:1–204, 2016) and physical constants are taken into account in the current calculation. The purpose of this work is two fold, one is to improve the earlier calculation of weak rates performed by Nabi and Klapdor-Kleingrothaus (At. Data Nucl. Data Tables 71:149, 1999a) using the same theory. We further compare our results with previous calculations. The selected \(sd\)-shell nuclei, considered in this work, are of special interest for the evolution of O–Ne–Mg core in 8–10 \(\mbox{M}_{\odot }\) stars due to competitive gamma ray heating rates and cooling by URCA processes. The outcome of these competitions is to determine, whether the stars end up as a white dwarf (Nabi in Phys. Rev. C 78(4):045801, 2008b), an electron-capture supernova (Jones et al. in Astrophys. J. 772(2):150, 2013) or Fe core-collapse supernova (Suzuki et al. in Astrophys. J. 817(2):163, 2016). The selected \(sd\)-shell nuclei for calculation of associated weak-interaction rates include \(^{20,23}\mathrm{O}\), \(^{20,23}\mathrm{F}\), \(^{20,23,24}\mathrm{Ne}\), \({}^{20,23\mbox{--}25}\mathrm{Na}\), and \({}^{23\mbox{--}25}\mathrm{Mg}\). The cooling and heating rates are calculated for density range (\(10 \leq \rho\ (\mbox{g}\,\mbox{cm}^{-3}) \leq 10^{11}\)) and temperature range (\(0.01\times 10^{9}\leq T(K)\leq 30\times 10^{9}\)). The calculated gamma heating rates are orders of magnitude bigger than the shell model rates (except for \(^{25}\mathrm{Mg}\) at low densities). At high temperatures the gamma heating rates are in reasonable agreement. The calculated cooling rates are up to an order of magnitude bigger for odd-A nuclei.

Gamma ray heating rates due to chromium isotopes in stellar core during late stages of high mass stars (>10M⊙)

Gamma ray heating rates are thought to play a crucial role during the pre-supernova stage of high mass stars. Gamma ray heating rates, due to β± -decay and electron (positron) capture on chromium isotopes, are calculated using proton-neutron quasiparticle random phase approximation theory. The electron capture significantly affects the lepton fraction (Ye ) and accelerates the core contraction. The gamma rays emitted as a result of weak processes heat the core and tend to hinder the cooling and contraction due to electron capture and neutrino emission. The emitted gamma rays tend to produce enormous entropy and set the convection to play its role at this stage. The gamma heating rates, on 50-60 Cr, are calculated for the density range 10 ) and temperature range 107 .

Potential Impacts of Modeling Full Reactor Cores Using Combined Fuel Performance and Thermal Hydraulics Codes

Nuclear reactor systems codes have the ability to model the system response in an accident scenario based on known initial conditions (ICs) at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermomechanical fuel rod response models needed for best-estimate prediction of fuel rod failure. Alternatively, the reverse can be said about fuel performance codes; they can lack the ability to capture and model the thermal-hydraulic (T-H) influence of adjacent fuel rods and the rod’s location in the reactor core. This work analyzes the limitations in using fuel performance codes to represent in-reactor conditions as determined by full-core T-H codes. The codes used in this analysis are the U.S. Nuclear Regulatory Commission’s steady-state fuel performance code FRAPCON-3.5 and T-H code TRACE-V5P3. In order to assess the impact of the limitations found in the codes, several modifications were made to all of the codes to improve code-to-code consistency. The modifications to the fuel performance code include adding the ability to model gamma-ray heating and providing realistic core coolant conditions. The T-H code modifications include adding the ability to model the fuel with axially varying burnup-dependent fuel and cladding dimensional changes and corrosion characteristics. The fuel in a Westinghouse four-loop pressurized water reactor was modeled to assess the impacts these modifications have on fuel performance and ICs for transient analysis. The results of this study show that current modeling assumptions (and limitations) can yield both conservative and nonconservative results on several important licensing criteria.

Effect of gamma-ray irradiation on structural properties of GaAsN films grown by metal organic vapor phase epitaxy

The effects of gamma-ray irradiation on the structural properties of GaAs1−xNx films (N concentration=1.9 and 5.1 at%) grown by metal organic vapor phase epitaxy on GaAs (001) substrates were investigated. The GaAs1−xNx films were irradiated by gamma rays with irradiation strength of 0–2.0MGy. Scanning electron microscopy and atomic force microscopy results showed that a gamma ray with a strength of 0, 0.5, 1.0, 1.5, and 2.0MGy formed holes with a density of 0.0, 8.8, 9.4, 11.5, and 11.9μm−2, respectively, on the surface of a GaAs0.981N0.019 film with low N content. On the other hand, the irradiated high-N-content GaAs0.949N0.051 film exhibited a cross-hatch pattern, which was induced by partial strain relaxation at high N levels, with a line density of 0.0, 0.21, 0.37, 0.67, and 0.26μm−1 corresponding to an irradiation strength of 0, 0.5, 1.0, 1.5, and 2.0MGy, respectively. The high-resolution X-ray diffraction and Raman scattering results revealed an increase in N incorporation and strain relaxation after irradiation. In addition, the GaAs0.949N0.051 films exhibited phase separation, which took place via N out-diffusion across the interface when the irradiation strength exceeded 1.0MGy. Based on these results, the main cause of structural change was determined to be the irradiation effects including displacement damage and gamma-ray heating.

Qualification of a gamma-ray heating calculation scheme for the future Jules Horowitz material testing reactor (RJH)

The objective of this study is to develop the modelling of the nuclear heating of the experimental devices of the future Jules Horowitz material testing reactor (RJH). The strong specific nuclear power produced (460 kW/l) induces intense photonic fluxes which cause heating and large gradients in temperature. The design studies of the experimental devices require an accurate knowledge of these gradients. A priori, we estimate that there are large uncertainties on gamma heating calculations to about 30% (2 σ). They are mainly due to the lack of knowledge about gamma emission yields and spectra and secondly come from the mismodelling of the gamma response function. In order to specify this uncertainty, a set of integral dose measurements using different kinds of TLD and a micro-ionization chamber are carried out in the UO 2 core of the MINERVE research reactor. The comparison between calculated and measured integral (prompt plus delayed) gamma-ray absorbed doses in TLD shows that calculation constantly underestimates the measurement with C/ E = 0.72 ± 15% ( k = 2) of total uncertainties. The systematic bias observed is partially attributed to a lack of gamma-ray emission data in basic nuclear data libraries used. A Bayesian method of adjustment is developed in order to re-estimate the principal components of the gamma heating and to transpose the results obtained to the devices of the RJH. This study made it possible to reduce significantly the uncertainties on the determination of the gamma heating from 30% to 15% (2 σ).

Thermal design study of a liquid hydrogen-cooled cold-neutron source

The use of both liquid hydrogen as a moderator and polycrystalline beryllium as a filter to enhance cold neutron flux at the UC Davis McClellan Nuclear Radiation Center has been studied. Although, more work is needed before an actual cold neutron source can be designed and built, the purpose of this preliminary study is to investigate the effects of liquid hydrogen and the thickness of a beryllium filter on the cold neutron flux generated. Liquid hydrogen is kept at 20 K, while the temperature of beryllium is assumed to be 77 K in this study. Results from Monte Carlo simulations show that adding a liquid hydrogen vessel around the beam tube can increase cold neutron flux by more than an order of magnitude. As the thickness of the liquid hydrogen layer increases up to about half an inch, the flux of cold neutrons also increases. Increasing the layer thickness to more than half an inch gives no significant enhancement of cold neutron flux. Although, the simulations show that the cold neutron flux is almost independent of the thickness of beryllium at 77 K, the fraction of cold neutrons does drop along the beam tube. This may be due to the fact that the beam tube is not shielded for neutrons coming directly from the reactor core. Further design studies are necessary for to achieve complete filtering of undesired neutrons. A simple comparison analysis based on heat transfer due to neutron scattering and gamma-ray heating shows that the beryllium filter has a larger rate of change of temperature and its temperature is higher. As a result heat will be transferred from beryllium to liquid hydrogen, so that keeping liquid hydrogen at the desired temperature will be the most important step in the cooling process.

Neutron and gamma ray heating in the grazing incident liquid metal mirrors for laser inertial fusion energy power plants

A thin film of liquid metal can serve as final optics of a laser inertial fusion energy (IFE) power plant. Calculations of pulsed neutron and gamma-ray heating are presented for a grazing incident liquid metal mirror (GILMM) used for robust final optics of a laser IFE power plant. Different liquid films (Li, Na, Mg, Al, Si, K, Ga, Ag, Au, Pb, Bi and Flibe “Li 2BeF 4”) are considered at a distance of 30 m from a nominal 1 GJ fusion source as well as different substrate materials (SS-304 and SiC). The effect of neutron heating both in the liquid metal film as well as in the subsrate material will be by around three to four orders of magnitude lower than the laser heating limit. Hence the nuclear heating will not be a limiting factor for grazing incident liquid metal mirror (GILMM) of a laser IFE power plant.

Benchmark experiment on silicon carbide with D–T neutrons and validation of nuclear data libraries

Silicon carbide (SiC) is one of the low activation structural materials for fusion reactors. A benchmark experiment on SiC was conducted at the deuterium–tritium (D–T) neutron source facility FNS in JAERI, and the validity of the cross section data for SiC was investigated. An experimental assembly made of sintered SiC (457×457 mm, and 711 mm in thickness) was bombarded by D–T neutrons, and neutron spectra, gamma-ray spectra, dosimetry reaction rates and gamma-ray heating rates were measured in the assembly. Computational analyses of the experiment were performed by Monte Carlo transport calculations with the MCNP-4B code and several recent evaluated nuclear data files. The following findings were pointed out. The calculations with JENDL Fusion File and JENDL-3.2 predicted adequately the neutron fluxes from 14 MeV to 0.3 eV as well as the gamma-ray fluxes at any positions up to ∼600 mm depth in the SiC assembly. Results for the calculation with FENDL/E-2.0 were similar to those with the two JENDL calculations, but the neutron fluxes in the energy range from 1 to 14 MeV were calculated slightly smaller. The cross section data for silicon in FENDL/E-1.0 were found inadequate for nuclear design calculations.

00/00784 On a simplified system for steam production in nuclear power plants with pressurized water reactors

The recently developed FENDL-1 database, both in multigroup form (FENDL/MG-1.0) and continuous energy form (FENDL/MC-1.0) has been tested through analyzing a fusion integral experiment performed at the FNS facility, Japan, on a large bulk shielding assembly made of multilayers of SS316 and water. The assembly is a replica that simulates ITER shielding blanket and is bombarded by a 14 MeV source placed at 30 cm from the cylindrical assembly and housed inside a SS316 cylindrical can. This activity is undertaken as part of co-operation with JAERI on executing the required neutronics R&D tasks for ITER shield design. The objectives are (a) benchmarking FENDL-1 data and identifying any flaws that may exist in this newly developed database, and (b) examining the range of discrepancy between the calculated nuclear parameters inside the assembly and the measured ones in terms of the ratio of the calculated-to-experimental (C/E) data. Both differential and integral experimental data were analyzed along the central axis of the ∼ 120 cm D × 140 cm L assembly. The analyses with the multigroup data, MG also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo (MC) code was used with the FENDL/MC-1 data. The largest range of discrepancy between calculated and measured responses (reaction rates, neutron spectra, gamma ray heating, etc.) was found to be ∼ 20–30% even though in most cases this discrepancy falls within the experimental errors.

Nuclear Analysis of Bulk Shielding Fusion Integral Experiments on Large SS316/Water Assembly with Simulated Super Conducting Magnet

As part of a collaboration with Japan, the U.S. participated in several fusion integral experiments that simulate the design features of the shielding blanket of the International Thermonuclear Experimental Reactor, ITER. The purpose of these efforts is to resolve the critical issues associated with the neutronics R&D tasks of ITER, among which is the adequacy of the newly developed FENDL-1 database. For that purpose, JAERI has constructed a cylindrical test assembly of dimension 1.2 D × 1.2 L m and made of front multi-layers of SS316 and water with an embedded smaller zone consists of multi-layers of super conducting magnet (SCM) stimulant and SS316. Measured parameters, covering the neutron energy range from 14 MeV down to thermal energy, were taken inside the SS316 and the SCM layers at 9 locations up to a depth of 91.4 cm. In one experiment (Assembly#l), a 1.27 cm B4C + 3.8 cm Pb layer was added in front of the SCM multi-layer zone. This layer is not included in Assembly#2. As in previous experiments, the 14 MeV source is housed inside a source reflector can (20 cm-thick) and located at a distance of 30 cm from the assembly. The U.S. analysis reported here was performed with 175n-42g FENDL/MG-1.0 (multigroup) as well as ENDF/B-VI data using the DORT 2-D code. Analysis was also performed with the Monte Carlo (MC) continuous energy data, FENDL/MC-1.0. The calculated parameters were compared to the following measured data: (a) neutron spectrum below 2 MeV, (b) foil activation rates such as Nb-93(n,2n)Nb-93m, Al-27(n,α)Na-24, In-115(n,n)In-115m, Au-197(n, γ)Au-198, and B-10(n,α)Li-7, (c) fission rate U-235(n,f) and U-238(n,f). (d) gamma-ray spectrum, and (e) gamma-ray heating rate.

Verification of ITER shielding capability and FENDL data benchmarking through analysis of bulk shielding experiment on large SS316/water assembly bombarded with 14MeV neutrons

Abstract The recently developed FENDL-1 database, both in multigroup form (FENDL/MG-1.0) and continuous energy form (FENDL/MC-1.0) has been tested through analyzing a fusion integral experiment performed at the FNS facility, Japan, on a large bulk shielding assembly made of multilayers of SS316 and water. The assembly is a replica that simulates ITER shielding blanket and is bombarded by a 14 MeV source placed at 30 cm from the cylindrical assembly and housed inside a SS316 cylindrical can. This activity is undertaken as part of co-operation with JAERI on executing the required neutronics R&D tasks for ITER shield design. The objectives are (a) benchmarking FENDL-1 data and identifying any flaws that may exist in this newly developed database, and (b) examining the range of discrepancy between the calculated nuclear parameters inside the assembly and the measured ones in terms of the ratio of the calculated-to-experimental (C/E) data. Both differential and integral experimental data were analyzed along the central axis of the ∼ 120 cm D × 140 cm L assembly. The analyses with the multigroup data, MG also included library derived from ENDF/B-VI data base for comparison purposes. The MCNP Monte Carlo (MC) code was used with the FENDL/MC-1 data. The largest range of discrepancy between calculated and measured responses (reaction rates, neutron spectra, gamma ray heating, etc.) was found to be ∼ 20–30% even though in most cases this discrepancy falls within the experimental errors.

Evaluation of the Neutron and Gamma-Ray Heating in the Radiation Shielding and Magnet Coils of the VISTA Spacecraft

Basic nuclear data for a design concept with inertial fusion energy propulsion for manned or heavy cargo deep space missions beyond earth orbit have been evaluated. Fusion power deposited in the inertial confined fuel pellet debris delivers the rocket propulsion with the help of a magnetic nozzle.The superconducting magnets of the magnetic nozzle are protected against neutron and gamma-ray radiation by a massive shielding. Throughout the shielding, the nuclear heating, caused by neutrons and gamma rays has been calculated. As a critical issue for this design concept, special attention is paid to the nuclear heating in the superconducting magnet coils. The neutron and gamma-ray penetration into the coils is calculated using the Sn methods with a high angular resolution in r-z geometry in S16 P3 approximation by dividing the solid space angle in 160 sectors.Total peak nuclear heat generation density in the coils is calculated as 64.5 μW/cm3 by a fusion power of 17 500 MW. Peak neutron heating density is 30.8 μW/cm3, and peak gamma-ray heating density is 40.6 μW/cm3 (on a different point). However, volume-averaged heat generation in the coils is much lower, namely, 2.17, 8.49, and 10.66 μW/cm3 for neutron, gamma-ray, and total nuclear heating, respectively.A conically shaped frozen hydrogen expellant reduces the neutron streaming toward the spacecraft by a factor of ~12.5 via neutron scattering on hydrogen and deflection into vacuum, in addition to the geometric neutron flux attenuation in space by 1/r2. The results of these calculations can help to increase the credibility of the vehicle for interplanetary space transport applications design concept.

Measurement of Gamma-Ray Spectra and Heating Rates in Iron and Stainless Steel Shields Bombarded by Deuterium-Tritium Neutrons and Validation of Secondary-Gamma-Ray Data in Evaluated Nuclear Data Libraries

Benchmark experiments for the validation of secondary-gamma-ray data are conducted for iron and Type 316 stainless steel (SS316) shield assemblies bombarded by deuterium-tritium neutrons. Gamma-ray spectra and heating rates for both threshold and capture gamma rays are measured. With the present experimental data for gamma rays, a set of benchmark data for iron and SS316, including neutron energy spectra in entire energies and various dosimetry reaction rates, is completed for the first time. Secondary-gamma-ray data in JENDL-3.1, JENDL-3.2, JENDL-Fusion File, and FENDL/E-1.0 are tested by benchmark calculation of the experiments. As a result, larger gamma-ray-production cross sections for threshold reactions in JENDL-3.1 and JENDL-3.2 and an inconsistent energy balance of the (n, γ) reactions in JENDL-3.1 are found. From the viewpoint of fusion engineering, the first priority in evaluating secondary-gamma-ray data should be conserving the energy balance. A rigid energy balance in both the JENDL Fusion File and FENDL/E-1.0 is confirmed for both threshold and capture gamma rays. The JENDL Fusion File and FENDL/E-1.0 provide the highly accurate secondary-gamma-ray data for iron and SS316 needed for fusion reactor nuclear design.

Measurement techniques for fusion blanket neutronics experiments

Abstract A variety of techniques to measure the nuclear parameters, such as tritium production rate, neutron spectrum, reaction rate and gamma-ray heating, in a simulated fusion blanket assembly have been developed or introduced through the JAERI/USDOE Collaborative Program on Fusion Blanket Neutronics. The features of those techniques are summarized and discussed with the experimental error. The present measurement techniques provided data with error ranges of 3–5% for tritium production, 5–10% for the neutron spectrum, 3–6% for the activation reaction and 10–20% for the gamma-ray heating rate.

Study of Gamma Radiation from Interaction of 14.7-MeV Neutrons with208Pb

The gamma radiation from the interaction of 14.7-MeV neutrons with {sup 208}Pb is investigated by high-resolution germanium-detector gamma-ray spectroscopy by using an enriched {sup 208}Pb sample. Cross sections for 14 gamma-ray lines from the {sup 208}Pb(n,n{prime}{gamma}) and {sup 208}Pb(n,2n{gamma}) reactions are measured at an emission angle of 124 deg. The results are compared with measurements from previous studies and with predictions based on the statistical theory of nuclear reactions (including direct and precompound contributions). The current results, especially for the {sup 208}Pb(n,n{prime}{gamma}) reaction, are considerably smaller than the results of most of the measurements of the previous studies probably because of the neglect of important sources of background, e.g., gamma-ray production in lead shielding, in the previous studies. Agreement with theory is adequate for the strong transitions between the lowest levels in {sup 207}Pb and {sup 208}Pb, but large discrepancies exist for the weaker transitions, especially for gamma-ray transitions from levels where experimental knowledge of branching ratios is missing. The study of the gamma radiation from the interaction of 14-MeV neutrons with lead is essential for accurate prediction of the gamma-ray heating in such devices. In addition, the nucleus {sup 208}Pb is especially studied as a test case for detailedmore » comparison of the predictions of nuclear reaction theory with experiments.« less

Benchmark experiment on copper with D-T neutrons for verification of secondary gamma-ray data in JENDL-3.1

Copper is a very important material for fusion devices because it is used as material for the first wall and the superconducting magnet. Nuclear heating of these components is a critical issue in the design of ITER. In order to verify secondary gamma-ray data in evaluated nuclear data libraries, a clean benchmark experiment on copper with D-T neutrons was carried out at the FNS facility. Gamma-ray spectra and heating rates were measured at several points in a experimental assembly of 629 mm in diameter and 608 mm in thickness along with neutron spectra and reaction rates. The experiment was analyzed by using MCNP-4 and DOT-3.5 with cross-section libraries based on JENDL-3.1. It is pointed out from the results that the secondary gamma-ray data for threshold reactions is good. The probability of direct transitions in JENDL-3.1 from the capture state to the ground state for keV neutron capture should be increased. It is also found that energy balance of secondary gamma-ray data for the neutron capture reaction is inconsistent in an incident neutron energy range between 0.01 and 10 keV. Excess of released gamma-ray energy in JENDL-3.1 results in larger gamma-ray heating rates.

Spectrum Weighting Function Method for In-Situ Fast Neutron and Gamma-Ray Response Measurements in Fusion Integral Experiments with an NE213 Scintillation Detector

Spectrum weighting function method with an NE213 scintillation detector has been applied to measurements of integral parameters such as tritium production rate from {sup 7}Li, in-system neutron integral flux and gamma-ray heating rate in fusion neutronics experiments. The NE213 scintillation detector can separate neutron and gamma-ray responses from each other, and those detector responses give energy information of both neutron and photon from recoil-proton and recoil-electron spectra, respectively. Connecting these energy responses with the energy responses of the nuclear parameters of interest such as {sup 7}Li(n,n{prime}{alpha}){sup 3}T reaction cross section, step function response of neutron energy and mass energy absorption coefficiency of gamma-ray, the corresponding nuclear parameters are obtained indirectly. The conversion method from the raw response of the detector is introduced by a spectrum weighting function.

Benchmark Experiment and Analysis of a Beryllium Cylindrical Assembly

AbstractAn integral experiment was performed on a Be cylindrical assembly of 630mm in diameter and 456mm in thick. Measured items were reaction rates, in-system neutron spectra and gamma-ray heating rates. The experimental analysis was performed by the MCNP and DOT3.5 codes using the nuclear data of JENDL-3, JENDL-3PR1, ENDF/B-IV and LANL. For high threshold reactions and integral flux above 10MeV, the calculation based on JENDL-3 agrees well with the experiment, while the calculations of ENDF/B-IV and LANL underestimetes those compared to the experiment. For integral neutron flux of 0.16∼0.5 MeV, the calculation of JENDL-3 agrees well with the experiment. But in the case of ENDF/B-IV the underestimation is 20%. It can be concluded that the nuclear data of Be in JENDL-3 improves very much in accuracy from the temporary version JENDL-3PR1.

Measurements of gamma-ray heating in lithium-oxide, graphite and iron slab assemblies bombarded by D—T neutrons

Measurements of the γ-ray heating rate distribution were made in simulated fusion blanket assemblies using thermo-luminescent dosimeters (TLD). The interpolation method, which needs to knowledge of the γ-ray spectrum, was applied using four kinds of TLDs ( 7 LiF, Mg 2 SiO 4 , Sr 2 SiO 4 , and Ba 2 SiO 4 ). To subtract the neutron contribution from the total dose, we calculated the neutron responses of these TLDs by a computer code based on the kerma calculations. Neutron responses of these TLDs were measured at two energy points using FNS at JAERI and FNG at ANL in order to examine the validity of the calculated results. The measurements were performed for lithium-oxide, graphite and iron slab assemblies bombarded by D—T neutrons. The overall experimental error was about 20 – 30%. It is suggested that development of new TLDs of low Z and low neutron sensitivity is desirable. The experimental results were compared with the calculated ones with the transport code DOT3.5 and two coupled neutron and γ-ray cross section sets GICX40 and GICXFNS to test the secondary γ-ray production cross section. It is shown that the accuracy of the calculational prediction of γ-ray heating is quite insufficient.

Measurement of gamma-ray heating with thermoluminescense dosimeters including Sr2SiO4 and Ba2SiO4.

Click to increase image sizeClick to decrease image sizeKEYWORDS: gamma-ray heatingthermoluminescense dosimetersfast critical assemblyJSD1000fast reactorsspatial distributionabsorbed dosesatomic numberSr2SiO4 Ba2SiO4