Determination Of Neutron Spectra Research Articles

Determination of neutron spectrum based on artificial neural network using liquid scintillation detector EJ-301.

This paper focuses on the neutron spectrum measurement using a liquid scintillation detector, where the neutron spectrum could be identified and unfolded from the light output distribution of the EJ-301 liquid scintillation detector through a linear artificial neural network (ANN). The response functions of the EJ-301 detector for monoenergetic neutron sources, as well as the light outputs, have been simulated and calculated by Monte Carlo procedure FLUKA. The linear ANN was trained and tested through the simulated data, where response functions were set as the input of ANN and the corresponding neutron spectra were output. Therefore, the neutron spectrum-unfolding model was created. This spectrum-unfolding model was tested through the light outputs induced by monoenergetic neutrons and the random superposition of them. Unfolding results show that this model could identify the information of the neutron spectrum accurately from the light outputs of a liquid scintillation detector. Moreover, the EJ-301 detector was used to measure the radioactivity of 252Cf, and the pulse height distribution induced by neutrons was derived through the charge-comparison method to remove the influence of gamma rays. The measured pulse height distribution was unfolded by the trained model, and measured results show that the unfolded neutron spectrum of 252Cf was consistent with the reference one. This paper presents the feasibility that the unknown neutron spectrum could be identified and confirmed through a linear neural network trained by simulated monoenergetic neutron response functions, which could be a candidate of choice for the determination of the neutron spectrum.

Determination of a Pu-Be source neutron spectrum at Czech Metrology Institute

Contrary to the radionuclide neutron sources of 252Cf and 241Am-Be type, which are most widely used for the dosimetric instrumentation calibrations, the spectral source strength and spectrum averaged fluence to dose equivalent conversion coefficients of Pu-Be source are not recommended by an international standard. This work describes the determination of those parameters for the Pu-Be source used at Czech Metrology Institute (CMI) by means of a Bonner spectrometer in combination with neutron emission rate measurement in a manganese bath.

Impact of initial MCNP spectrum guess on experiment-based neutron spectrum determination at Missouri S&T reactor

The energy spectrum of the prompt-neutron flux at the Missouri University of Science and Technology Reactor (MSTR) was obtained using an activation method. Foils were irradiated at the bare rabbit tube (BRT) of the reactor core (which has a 120 W configuration). The neutron spectrum was determined using an unfolding method implemented in SAND-II code. The Monte Carlo N-particle (MCNP) model was used to calculate the spectrum at the bare rabbit tube, which was used as the initial guess for input into the SAND-II code. Various MCNP spectra with 620-, 143-, 89-, 50-, 22-, and 12-energy groups were used as the initial guess for input into SAND-II. Seventeen different foils were irradiated at 100 kW. The foil set covers energies from 0.025 eV to 13 MeV for a broad spectrum analysis. The photon counts for the activated foils were obtained using a high-purity germanium (HPGe) detector. The count results were input into the SAND-II code to predict the neutron flux spectrum for the MSTR. The spectra were collapsed into three groups: thermal, epithermal, and fast flux. With the 620-group initial guess, the thermal, epithermal, and fast neutron fluxes were 1.43×1012±2.82×1011n/cm2s, 4.51×1011±2.85×1010n/cm2s, and 5.38×1011±4.85×1009n/cm2s, respectively, giving a total flux 2.42×1012±3.02×1011n/cm2s. Disparities were noted in the distribution of the thermal and epithermal flux predictions, as the number of groups in the initial spectrum guess changed. The 59%/19% distribution of the thermal/epithermal flux, as predicted with the 620-group guess, is inconsistent with the 49%/26% distribution predicted with the 89- and 143-group guesses. The predictions based on the 89- and 143-group guesses are fairly consistent with the predictions obtained with the 22- and 50-group guesses. The 12-group initial guess resulted in the prediction of a fairly even distribution between the thermal (38%) and epithermal (36%) fluxes. Regardless of the number of groups in the initial guess, the SAND-II prediction is fairly consistent in the fast energy range. The fast neutron flux was found to range between 22% and 26%.

Neutron spectrum determination of p+Be reaction for 30 MeV protons using the multi-foil activation technique

At the NPI in Rez, the p + Be source reaction was investigated for 30 MeV proton beam and thick beryllium target. For neutron field determination of the p(30)+Be source reaction in close source-to-sample distance, the multi-foil activation technique with a set of 10 activation materials (Au, Co, Lu, Ti, In, Al, Y, Fe, Ni, Nb) was utilized. From resulting reaction rates, the neutron spectrum was reconstructed using the SAND-II unfolding code. New neutron field of white spectrum up to 28 MeV has an intensity of 8.6 × 1010 cm−2s−1 close to target. The obtained neutron field extends the utilization of cyclotron-based fast neutron sources at the NPI and provides new experimental opportunities for future irradiation experiments such as fast neutron activation analysis, nuclear data validation, and radiation damage study of electronics and materials for nuclear energetics.

Numeric Determination and Validation of Neutron-Induced Radioactive Nuclide Inventories for Decommissioning and Dismantling of Light Water Reactors

The decommissioning of nuclear power plants requires project planning and budgeting both during the project and in advance, as well as the secured provision of financial and human resources. When a facility is free from irradiated fuel, the reactor pressure vessel with the nuclear components and the biological shield determine the activity inventory of the facility, which almost exclusively consists of activated radionuclides located in the respective structures. Knowledge of the activity distribution and nuclide vectors of the involved components is of vital importance for decommissioning planning. In this context, the development of a computation procedure is described coupling the Monte Carlo method for the determination of neutron flux densities and spectra with a procedure to perform activation calculations for the determination of nuclide vectors. For this purpose, detailed knowledge of the material composition, particularly the trace-element concentrations of nitrogen and cobalt in steel and additionally of europium and cesium in concrete structures, considerably impacts the accuracy of the calculated activities. Extensive validation using data collected from various nuclear power plants to be decommissioned, such as nuclide activities, neutron flux densities, and neutron and gamma dose rates, demonstrates the reliability of the computed nuclide distributions showing ratios of computed-over-measured values of typically between 0.9 and 3. The practicality of the developed method and the convenient use of the results have already been demonstrated analyzing several German boiling water reactors and pressurized water reactors and developing packaging strategies based on the produced results.

Neutron spectrum determination of d(20)+Be source reaction by the dosimetry foils method

The cyclotron-based fast neutron generator with the thick beryllium target operated at the NPI Rez Fast Neutron Facility is primarily designed for the fast neutron production in the p+Be source reaction at 35MeV. Besides the proton beam, the isochronous cyclotron U-120M at the NPI provides the deuterons in the energy range of 10–20MeV. The experiments for neutron field investigation from the deuteron bombardment of thick beryllium target at 20MeV were performed just recently. For the neutron spectrum measurement of the d(20)+Be source reaction, the dosimetry foils activation method was utilized. Neutron spectrum reconstruction from resulting reaction rates was performed using the SAND-II unfolding code and neutron cross-sections from the EAF-2010 nuclear data library. Obtained high-flux white neutron field from the d(20)+Be source is useful for the intensive irradiation experiments and cross-section data validation.

Determination of neutron spectrum outside the lateral shielding of Taiwan Photon Source using high-sensitivity Bonner cylinders

A homemade Bonner cylinder spectrometer was used to determine the neutron spectrum from thermal energy up to several GeV at a heavily-shielded light source facility. The spectrometer, similar to the design of Bonner spheres, features high sensitivity of neutron detection because of using a long cylindrical 3 He proportional counter. The spectrum measurement was performed during the facility commissioning by intentionally parking the injected electrons at the septum of the storage ring. Based on a high-fidelity FLUKA simulation, the predicted neutron spectrum at the location of measurement was adopted as the initial guess of spectrum unfolding. The unfolded result indicated an underestimation of the calculated neutron spectrum in the high-energy portion, leading to a substantial revision of the neutron dose rate at the location.

ACTIVATION ANALYSIS OF HISTORICAL SAMPLES AND NEUTRON SPECTRUM DETERMINATION AT VR-1 TRAINING REACTOR

Vertical irradiation channel of the VR-1 training reactor of the Department of Nuclear Reactors CTU was used for activation analysis of historical samples from the 14th to the 19th century. For determination of mass fractions of materials such as copper, gold and silver in irradiated samples, the relative method of activation analysis was used. Further, a set of 24 activation detectors of 12 various materials served for the determination of neutron spectrum of the VR-1 reactor using analytical method; moreover the analytical solution was compared to unfolded spectrum obtained from SAND-II deconvolution program.

Penentuan Spektrum Neutron di LINAC Menggunakan Passive Single Sphere Spectrometer

Neutron spectra at Linear Acceleration (LINAC) facility was needed to protect neutron radiation. The purpose of determination of neutron spectra at LINAC using Passive Single Sphere Spectrometer is getting spectra and neutron fluence from LINAC and knowing the influence of input voltage at LINAC to its spectra and neutron fluence output. Single Sphere Spectrometer (SSS) with diameters 30 cm using seven gold foil detectors was set in position a half of diameter (0, 3, 6, 9, 11, 12, 13 cm) from center of the SSS. Measurement of neutron count rate was performed by placing SSS under gantry of LINAC with SSD 100 cm in distances. Next, neutron count rate was used as input data in UMG 3.3 program to get spectra. Based on measurement, total of neutron fluence at LINAC 10 MV and 15 MV is 5,78 ´ 10 4 n/cm 2 .s ± 0,01 ´ 10 4 and 9,44 ´ 10 5 n/cm 2 .s ± 0,22 ´ 10 5 , so increasing input voltage at LINAC was followed by increased neutron fluence.

Verification of a fusion neutron diagnostic Bonner sphere spectrometer on measurement of a 241Am–Be neutron source

A Bonner sphere spectrometer (BSS) was developed for neutron diagnostic on HL_2A Tokamak. It contains eight polyethylene spheres embedded with SP9 3He proportional counter. Before setting up on the Tokamak experimental hall, a verification experiment was arranged on a 241Am–Be neutron source to test its spectrometry capability. The neutron response functions were calculated by Monte Carlo code Geant4 to simulate the real measurement environments. By least square method, the neutron spectrum was finally unfolded on log domain from 0.1 eV to 11 MeV. It has a remarkable consistency to the ISO 8529-1 standard 241Am–Be neutron spectrum. This shows that the BSS is effective and reliable for neutron spectrum determination.

Determination of Neutron Spectra Within the Energy of 1 keV to 1 MeV by Means of Reactor Dosimetry

The standard procedure for neutron reactor dosimetry is based on neutron irradiation of a target and its post-irradiation analysis by Gamma and/or X-ray spectrometry. Nowadays, the neutron spectra can be easily characterized for thermal and fast energies (respectively 0.025 eV and $ > 1~\hbox{MeV}$ ). In this paper we propose a new target and an innovating post-irradiation technique of analysis in order to detect the neutron spectra within the energy of 1 keV to 1 MeV. This paper will present the calculations performed for the selection of a suitable nuclear reaction and isotope, the results predicted by simulations, the irradiation campaign that is proposed, and the post-irradiation technique of analysis.

Neutron Spectrum Determination of the p(35 MeV)-Be Source Reaction by the Dosimetry Foils Method

The thick target neutron field of source reaction p + Be was investigated for a proton energy of 35 MeV. The spectral neutron flux at 0∘ for two target-to-sample distances was determined by using the dosimetry foils activation method. The present p(35)-Be white neutron spectra provide the suitable basis for irradiation experiments and integral tests of nuclear data.

Analysis of the application of the shadow cone technique for the determination of neutron spectrum with Bonner sphere spectrometer

In this study, shadow cones were examined with Bonner sphere spectrometer (BSS) for the measurement of neutron spectrum of an 241Am–Be source. The BSS consists of a set of nine polyethylene spheres with a central 6LiI(Eu) scintillator sensitive to thermal neutrons. The energy response matrix of the BSS was determined by Monte Carlo simulations for the neutron energies from 10−8 to 100MeV. A crucial problem in neutron measurements is determining the contribution of scattered neutrons coming from the structures and materials surrounding a source. For this purpose two shadow cones were manufactured and positioned between the source and the detector. Experiments were performed to select the optimal source-to-shadow cone distance. An unfolding code was used to obtain the neutron spectrum. The difference between the unfolded spectrum obtained by using the shadow cones and the ISO reference spectra was considered as a parameter to select the optimum condition in the shadow cone technique application.

Shielding Experiments under JASMIN Collaboration at Fermilab (IV) Measurement and Analyses of High-Energy Neutron Spectra in the Anti-Proton Target Station

Neutron spectra in high-energy region between 1 and 100 MeV in the shield configuration of the anti-proton target station and a 120-GeV proton beam at Fermi National Accelerator Laboratory (Fermilab) were determined using the reaction rate data obtained with the multi-foil activation method. Two kinds of methods were employed for the determination of neutron spectra: one is the fitting method which is newly developed in this work, another is the unfolding method with SAND-II code. The calculations were performed using the PHITS. From the comparison between the calculated and experimental neutron spectra, it concluded that the PHITS can be used for shielding design of highenergy proton accelerators.

Self-shielding factor calculations of heterogeneous samples in activation measurements for neutron spectrum unfolding

• Activation measurements for samples irradiated in CEA EOLE reactor central channel. • Monte Carlo (2 codes) calculated spectra and self-shielding factors in samples. • The C/M agreement is mostly within the uncertainty of up to 10%. • Construction of the spectrum covariance matrix from analytical function parameters. • Adjustment of the calculated spectrum by taking into account the measurements. Activation measurements can be used for neutron spectrum determination in irradiation facilities. It is very important that self-shielding is taken correctly into account. Activation measurements of monitor materials were performed at the CEA EOLE facility. Self-shielding factors for the monitor materials were calculated at the Commissariat a l’Energie Atomique with the TRIPOLI 4.6 code using a detailed whole-core model, and at the Jožef Stefan Institute with the MCNP code using a simplified model. The results were compared to assess the relative merits and the achievable accuracy of the calculations. In the end, the results were used to determine the neutron spectrum of the studied location in the EOLE irradiation facility.

Determination of workplace neutron spectra at a high energy hadron accelerator using active and passive Bonner sphere spectrometers

In the framework of the 2006 experimental benchmark organized at the GSI (Darmstadt, Germany) by the EC CONRAD network, a neutron dosimetry intercomparison was performed in a workplace field around a carbon target hit by 400 MeV/u C 12 ions. The radiation protection group of the INFN-LNF participated to the intercomparison with a Bonner sphere spectrometer equipped with an active LiI ( Eu ) 6 scintillator and a set of passive detectors, namely MCP-6s ( 80 mg cm - 2 ) /MCP-7 TLD pairs from TLD Poland. Both active and passive spectrometers, independently tested and calibrated, were used to determine the field and dosimetric quantities in the measurement point. The FRUIT unfolding code, developed at the INFN-LNF radiation protection group, was used to unfold the raw BSS data. This paper compares the results of the active or passive spectrometers, obtaining a satisfactory agreement in terms of both spectrum shape and value of the integral quantities, as the neutron fluence or the ambient dose equivalent. These results allow qualifying the BSS based on TLD pairs as a reliable passive method to be used around high energy particle accelerators even in low dose rate areas. This is particularly useful in those workplaces where the active instruments could be disturbed by the presence of pulsed fields, large photon fluence or electromagnetic noise.

Development of activation foils method for the IFMIF neutron flux characterization

This work proposes to use the activation foil method for the neutron spectrum determination in the IFMIF facility. Suitable elements were selected to this end with dosimetry reactions having thresholds and effective cross sections covering the whole IFMIF energy range up to 55 MeV. The available evaluated nuclear data libraries and experimental data were carefully screened to select appropriate dosimetry cross sections with uncertainties as low as possible. For the determination of the neutron spectra from the induced γ-activities the unfolding code SAND-II was applied after having it modified for the use of the selected dosimetry cross sections above 20 MeV. This approach was tested by measuring the white neutron spectrum at the cyclotron of Nuclear Physics Institute (Řež), where neutrons up to 35 MeV energy were generated by a 37 MeV proton beam impinging on a heavy water target. The paper concludes with recommendations for the procedure for determining the neutron energy differential flux in the IFMIF test cell with respect to the size of the dosimetry foils, the irradiation time and the expected induced activities.

Bonner sphere neutron spectrometry at nuclear workplaces in the framework of the EVIDOS project

The Institute for Radiological protection and Nuclear Safety was engaged in the EC funded EVIDOS project to provide reference spectrometry data using its Bonner sphere system. The data were processed by means of two unfolding codes, NUBAY and GRAVEL, both provided by the Physikalisch-Technische Bundesanstalt. The NUBAY program, based on Bayesian parameter estimation methods, assumes a parameterised spectrum and provides posterior probability distributions for the parameters. The code GRAVEL, an iterative algorithm based on SAND-II, was used with various default spectra, among them the NUBAY solution. The BS measurements were used to establish the neutron fluence energy distributions and reference values for the neutron ambient dose equivalent. As this quantity depends strongly on the high energy neutrons, a sensitivity analysis was done by unfolding the BS data with GRAVEL using the NUBAY solution spectrum as default with various changes in the parameters of the high energy peak. This new method of analysing Bonner sphere data allowed the determination of reliable neutron spectra, as well as a very good estimate of the corresponding integral quantities with small associated uncertainties.

MO-D-T-617-03: Secondary Neutron Spectra From High Energy IMRT and Conventional Treatment Plans Using Bonner Spheres and Au-197 Activation Foils

Purpose: The purpose of this work is to measure and compare secondary neutron spectra both in‐field and out‐of‐field from the delivery of high energyIMRT and conventional radiation therapy. Direct measurement of neutron spectra in phantom for IMRT has received little attention. Studies are needed to demonstrate the strength of the assumption that the spectra are the same for IMRT and conventional radiation therapy.Method and Materials: 197 Au activation foils are placed on the surface of a special holder which is inserted into different Bonner Spheres. Seven levels of moderation were used: the bare foil holder and the holder inside 2″, 3″, 5″, 8″, 10″, and 12″ spheres. Net counts of the 411keV gamma peak were measured for each foil using a high purity Germanium detector with spectral analysis capabilities. The activity at the end of irradiation, and the production rate per unit mass of target (Bqs−1g−1) were calculated. Data were unfolded with MXD_FC33 algorithm, PTB (Germany) with a response matrix specifically calculated for this measurement system using MCNPX. The response matrix was verified by unfolding 197 Au ‐Foil Bonner Sphere system data from irradiations with 252 Cf . Results: The unfolded neutron spectra and neutron flux are evaluated. Ambient Dose Equivalent, H*(10) is calculated from spectral data. The data show an increase in the secondary neutron flux and H*(10) for IMRT compared to conventional radiotherapy. The IMRT spectra show a small shift to lower energies. Ratios of in‐field to out‐of‐field flux for 18MV fall off approximately with the inverse square. The ratios of 18MV to 15MV IMRT and conventional out of field flux are 2.13 and 2.6, respectively. Conclusion: 197 Au activation foils inside Bonner spheres technique for the measurement of neutron spectra inside the treatment room is an effective tool for the determination of neutron spectra from IMRT and conventional radiation delivery.

Neural Networks for the Neutron Spectrum Determination Based on the Foil Activation Method

The multi-foil activation method using (n,p) and (n,α) of threshold reactions coupled with the neural network technique is considered to determine the neutron spectra of the 252Cf fission neutron source of the Research Institute for Radiation Biology and Medicine (RIRBM), Hiroshima University and a reactor at Kinki University (UTR-KINKI). Training data for the neural network were prepared from computational data sets of Monte Carlo calculation. The reproducibility of the calculated activities was verified to be within 15% of the experimentally determined activities. As it has been shown that the incident neutron spectrum of the 252Cf fission neutron source will reproduce the Maxwellian function, the reactor neutron spectrum of UTR-KINKI is successfully determined. A comparison with a previous evaluation of the reactor spectrum using an unfolding NEUPAC code is also given.