Neutron Source Strength Research Articles

A Compton-suppression detection system for use in manganese bath measurements

The manganese sulfate bath technique is a standard tool for neutron source strength measurement (Park et al., 2005). However, the dominate Compton continuum of most sodium iodide scintillators used in manganese bath systems (MBSs) does not allow the precise identification of induced gamma rays required for such measurements. In this research, to resolve this problem, a Compton-suppression system has been proposed which consists of a 2 in. by 2 in. NaI(Tl) right cylindrical scintillator as the main and a set of eight rectangular NE102 plastic scintillators of 12×12×15 cm3 dimensions as suppression detectors. Both detectors operate in anti-coincidence circuit to suppress the Compton continuum. The proposed system has been simulated with the MCNPX code with two different approaches and the corresponding measurements with 137Cs gamma-ray source and neutron-activated MnSO4 solution have been undertaken that give rise to a promising agreement.

Advancements in the development of a directional-position sensing fast neutron detector using acoustically tensioned metastable fluids

Advancements in the development of a direction and position sensing fast neutron detector which utilizes the directional acoustic tensioned metastable fluid detector (D-ATMFD) are described. The resulting D-ATMFD sensor is capable of determining the direction of neutron radiation with a single compact detector versus use of arrays of detectors in conventional directional systems. Directional neutron detection and source positioning offer enhanced detection speeds in comparison to traditional proximity searching; including enabling determination of the neutron source shape, size, and strength in near real time. This paper discusses advancements that provide the accuracy and precision of ascertaining directionality and source localization information utilizing enhanced signal processing-cum-signal analysis, refined computational algorithms, and on-demand enlargement capability of the detector sensitive volume. These advancements were accomplished utilizing experimentation and theoretical modeling. Benchmarking and qualifications studies were successfully conducted with random and fission based special nuclear material (SNM) neutron sources (239Pu–Be and 252Cf). These results of assessments have indicated that the D-ATMFD compares well in technical performance with banks of competing directional fast neutron detector technologies under development worldwide, but it does so with a single detector unit, an unlimited field of view, and at a significant reduction in both cost and size while remaining completely blind to common background (e.g., beta-gamma) radiation. Rapid and direct SNM neutron source imaging with two D-ATMFD sensors was experimentally demonstrated, and furthermore, validated via multidimensional nuclear particle transport simulations utilizing MCNP-PoliMi. Characterization of a scaled D-ATMFD based radiation portal monitor (RPM) as a cost-effective and efficient 3He sensor replacement was performed utilizing MCNP-PoliMi simulations, the results of which are discussed and presented.

Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz.

In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations. The depletion calculation code origen2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition of the current core was used in a mcnp5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established attila model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms. The mcnp5 simulated neutron spectrum and source strength are found to be in good agreement with the previous attila model whereas the photon production is much lower. Both mcnp5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom. A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the mcnp5 simulations and experiments demonstrates that the attila model overestimates the gamma dose contribution. The detailed model can be used for the planning of structural modifications in the thermal column irradiation channel or the use of different irradiation sites than the thermal column, e.g., the beam tubes.

SU‐F‐BRE‐11: Neutron Measurements Around the Varian TrueBeam Linac

Purpose:With the emergence of flattening filter free (FFF) photon beams, several authors have noted many advantages to their use. One such advantage is the decrease in neutron production by photonuclear reactions in the linac head. In the present work we investigate the reduction in neutrons from a Varian TrueBeam linac using the Nested Neutron Spectrometer (NNS, Detec). The neutron spectrum, total fluence and source strength were measured and compared for 10 MV with and without flattening filter and the effect of moderation by the room and maze was studied for the 15 MV beam.Methods:The NNS, similar to traditional Bonner sphere detectors but operated in current mode, was used to measure the neutron fluence and spectrum. The NNS was validated for use in high dose rate environments using Monte Carlo simulations and calibrated at NIST and NRC Canada. Measurements were performed at several positions within the treatment room and maze with the linac jaws closed to maximize neutron production.Results:The measurements showed a total fluence reduction between 35‐40% in the room and maze when the flattening filter was removed. The neutron source strength Qn was calculated from in‐room fluence measurements and was found to be 0.042 × 102 n/Gy, 0.026 × 102 n/Gy and 0.59 × 1012 n/Gy for the 10 MV, the 10 MV FFF and 15 MV beams, respectively. We measured ambient equivalent doses of 11 mSv/hr, 7 mSv/hr and 218 mSv/hr for the 10 MV, 10 MV FFF and 15 MV by the head.Conclusion:Our measurements revealed a decrease in total fluence, neutron source strength and equivalent dose of approximately 35‐40% across the treatment room for the FFF compared to FF modes. This demonstrates, as expected, that the flattening filter is a major component of the neutron production for the TrueBeam.The authors greatly acknowledge support form the Canadian Nuclear Commission and the Natural Sciences and Engineering Research Council of Canada through the CREATE program. Co‐authors Dubeau and Witharana are employees of Detec (Gatineau, Quebec), manufacturer of the Nested Neutron Spectrometer.

Robust reactivity, neutron source, and precursor estimators for nuclear reactors

Reactivity, precursor concentration, and external neutron source strength determine control, operation and performance of nuclear reactors. These main reactor quantities are not directly measurable and must be calculated or estimated using reactor kinetics. This study presents efficient and robust nonlinear estimation algorithms for predicting these fundamental reactor quantities. The effectiveness of the proposed estimators is assessed through chirp and step test signals in the presence of parameter uncertainties and measurement noise.

Evaluation of an alternative convenient irradiation system for determination of emission rate of radio-isotopic neutron sources

In present work, an alternative irradiation system based on a symmetric cylindrical tank filled with a moderator containing hydrogen, which was equipped with a NaI(Tl) scintillation detector, was proposed for using in determination of neutron flux. This irradiation system was designed by MCNP4C code, with considering a 241Am–Be neutron source in several volumes and different materials. When the neutron is captured by hydrogen, a 2.22 MeV prompt gamma-ray is emitted. The gamma pulse-height spectrum shows a photo-peak around 2.22 MeV whose net area is proportional to the total emission rate of neutron. The simulation result showed that a cylindrical tank with 110 cm diameter and height filled with water can be a suitable system for neutron source strength calibration. Furthermore, a proper two-layer shielding must be placed between the source and detector for preventing neutrons and gamma rays to directly enter the detector.

Monte Carlo characterizations mapping of the (γ,n) and (n,γ) photonuclear reactions in the high energy X-ray radiation therapy

AimThe aim of this work was to map the characteristics of (n,γ) and (γ,n) reactions in a high energy photon radiation therapy. BackgroundPhotoneutrons produced in the high energy X-Ray radiation therapy may damage patients and staff. It is due to high RBE of the produced neutrons according to their energy and isotropic emission. Characterization of the photoneutrons can help us in appropriate shielding. Materials and methodsThis study focused on the photoneutron and capture gamma ray phenomena. Characteristics such as dose value, fluence and spectra of both the neutrons and the by produced prompt gamma ray were described. Results and discussionNeutron and prompt gamma spectra in different points showed the neutrons to be thermalized when increasing the distance from the linac. Energy of the neutrons changed from about 0.6MeV at the isocentre to around 10−08MeV at the outer door position. Although the neutrons were found as fast neutrons, their spectra showed they were thermal neutrons at the outer door position. Additionally, it was seen that the energy of the gamma rays is higher than the scattered X-ray energy. The energy of gamma rays was seen to be up to 10MeV while the linac photons had energy lower than 1MeV. Neutron source strength obtained in this work was in good agreement with the published data, which may be a confirmation of our simulation accuracy. ConclusionThe study showed that the Monte Carlo simulation can be applied in the radiotherapy and industrial radiation works as a useful and precise estimator. We also concluded that the dose from the prompt gamma ray at the outer door location is higher than the scattered radiation from the linac and should be considered in the shielding.

Determining 252Cf source strength by absolute passive neutron correlation counting

Physically small, lightly encapsulated, radionuclide sources containing 252Cf are widely used for a vast variety of industrial, medical, educational and research applications requiring a convenient source of neutrons. For many quantitative applications, such as detector efficiency calibrations, the absolute strength of the neutron emission is needed. In this work we show how, by using a neutron multiplicity counter the neutron emission rate can be obtained with high accuracy. This provides an independent and alternative way to create reference sources in-house for laboratories such as ours engaged in international safeguards metrology. The method makes use of the unique and well known properties of the 252Cf spontaneous fission system and applies advanced neutron correlation counting methods. We lay out the foundation of the method and demonstrate it experimentally. We show that accuracy comparable to the best methods currently used by national bodies to certify neutron source strengths is possible.

Numerical analysis of startup PWR with fractional neutron point kinetic equation

In this paper we present the numerical analysis of the neutron density behavior when the nuclear reactor power is increased during startup of a PWR. The fractional neutron point kinetic (FNPK) equation with one-group delayed neutron precursor and external neutron source was used for this analysis. It is considered that there is a relaxation time associated with a rapid variation in the neutron flux and this effect is considered with the FNPK which have a physical interpretation of the fractional order is related with the sub-diffusive process, i.e., non-Fickian effects from the neutron diffusion equation point of view. In order to study the relaxation time effects during start-up of a PWR, a numerical analysis with FNPK is carried out, which it is assumed that during the ith step of control rod withdrawal the way of reactivity insertion is step to step, where the neutron source strength was defined as a constant in terms of a known initial stable sub-criticality and the neutron signal from a steady state condition. The results of the FNPK were compared with the classical neutron point kinetics (CNPK), for different values of the anomalous relaxation time.

Diagnostic neutron activation system for KSTAR

The Korea Superconducting Tokamak Advanced Research (KSTAR) device has begun deuterium plasma operation and increased neutron generation from D-D fusion reaction in the plasma is expected as the heating power of the plasma increases. A neutron activation system utilizing the pneumatic transfer of encapsulated metal samples has been implemented to monitor the fusion neutron source strength and the total fusion power from the KSTAR plasma. The prototype pneumatic transfer system for the ITER neutron activation system was slightly modified to be used for KSTAR, and a High Purity Germanium (HPGe) detection system was used to count gamma photons from the activated samples. The Monte Carlo code MCNP and the inventory code FISPACT were also used for the calculations to evaluate the total number of neutrons emitted from the D-D fusion reactions in the KSTAR plasma. The analysis of data from an activation measurement obtained from the 2011 KSTAR campaign shows the neutron flux at the irradiation station was 2.2 × 108 cm−2s−1, and the total neutron yield was 4.7 × 1013 n/s for a typical NBI-heated, H-mode KSTAR plasma shot.

Production and characterization of a custom-made 228Th source with reduced neutron source strength for the Borexino experiment

A custom-made 228Th source of several MBq activities was produced for the Borexino experiment to study the external background of the detector. The aim was to reduce the unwanted neutron emission produced via (α,n) reactions in ceramics typically used for commercial 228Th sources. For this purpose a ThCl4 solution was chemically converted into ThO2 and embedded in a gold foil.The paper describes the production of the custom-made source and its characterization by means of γ-activity, dose rate and neutron source strength measurements. From γ-spectroscopic measurements it was deduced that activity transfer from the initial solution to the final source was >91% (at 68% C.L.) and the final activity was (5.41±0.30) MBq. The dose rate was measured with two dosimeters yielding 12.1mSv/h and 14.3mSv/h in 1cm distance. The neutron source strength of the 5.41MBq 228Th source was determined to be (6.59±0.85) s−1.

Computation of Neutron Multiplicity Statistics Using Deterministic Transport

Nuclear nonproliferation efforts are supported by measurements that are capable of rapidly characterizing special nuclear materials (SNM). Neutron multiplicity counting is frequently used to estimate properties of SNM, including neutron source strength, multiplication, and generation time. Different classes of models have been used to estimate these and other properties from the measured neutron counting distribution and its statistics. This paper describes a technique to compute statistics of the neutron counting distribution using deterministic neutron transport models. This approach can be applied to rapidly analyze neutron multiplicity counting measurements without relying on the point reactor kinetics model.

Neutron diagnostics at the Wendelstein 7-X stellarator

The stellarator W7-X, presently under construction at the Institute for Plasma Physics in Greifswald, will be equipped with a set of neutron monitors in order to study the time behaviour of neutron emission generated during D-D plasma operation and neutral beam heating with Deuterium. Each of these neutron monitors consists of several neutron detector tubes inserted in a dedicated moderator. The neutron monitors at W7-X are designed to monitor neutron yields with a time resolution of 5 ms and with a statistical uncertainty of better than 15%. One of the monitors is located in the centre above the stellarator. The other five monitors are distributed around the torus. A prerequisite for the determination of the absolute neutron source strength produced by D(d,n)3He fusion reactions in the plasma is an in-situ calibration with a neutron source of known source strength. During such a calibration procedure, the neutron source will be moved along the torus axis and the count rates of the different neutron monitors will be measured. In a first benchmark experiment, an 241AmBe neutron source was moved along the torus axis within one module of the stellarator and the neutron signals were measured by a De Pangher Long Counter outside of the cryostat chamber as a function of the neutron source position. These measurements have been compared with predictions of Monte Carlo calculations (MCNP) of the neutron propagation from the location of the neutron source to the long counter. The concept of neutron monitors will be reported together with results from the benchmark experiment and results from MCNP calculations. The neutron monitor system is the first part of several neutron diagnostic systems such as neutron activation system, neutron profile camera planned for future neutron analysis. A short survey of these neutron diagnostic tools of W7-X will be given.

Fusion-Fission Hybrid Using a D-D Cylindrical Inertial Electrostatic Confinement (IEC) Driver

Recent progress in fusion development combined with the rebirth of nuclear fission power has regenerated interest in fusion-fission hybrid reactors. Such systems could be applied to both low power research reactors for use in University and industrial research assemblies and power reactors. However most attention has been directed at D-T fusion drivers using Tokamak, ICF or various alternate confinement systems like FRCs. However, the necessity to have large devices and breed tritium in the blanket complicates the concept. Here we propose the inertial electrostatic confinement (IEC) fusion approach since it offers the advantages of simple structural, high power density and a non-Maxwellian beam dominated plasma suited for burning advanced fuels to minimize tritium involvement. The cylindrical IEC allows a small compact unit which can be inserted into fuel element slots in the fission reactor core, thus providing a compact overall system and excellent neutronic coupling. The basic physics for the IEC has been demonstrated in small-scale laboratory experiments close to levels needed for driving a subcritical assembly for use in student teaching labs. However, for use in future high power hybrids significant scale-up in source strength is required. Scale up using an external ion source (e.g. a Helicon) so the background gas pressure is minimized in the reaction zone potentially offers a route to the required neutron source strength.

Dosimetry and fast neutron energies characterization of photoneutrons produced in some medical linear accelerators

This work focusses on the estimation of induced photoneutrons energy, fluence, and strength using nuclear track detector (NTD) (CR-39). Photoneutron energy was estimated for three different linear accelerators, LINACs as an example for the commonly used accelerators. For high-energy linear accelerators, neutrons are produced as a consequence of photonuclear reactions in the target nuclei, accelerator head, field-flattening filters and beam collimators, and other irradiated objects. NTD (CR-39) is used to evaluate energy and fluence of the fast neutron. Track length is used to estimate fast photoneutrons energy for linear accelerators (Elekta 10 MV, Elekta 15 MV, and Varian 15 MV). Results show that the estimated neutron energies for the three chosen examples of LINACs reveals neutron energies in the range of 1–2 MeV for 10 and 15 MV X-ray beams. The fluence of neutrons at the isocenter (Φtotal) is found to be (4×106 n cm2 Gy−1) for Elekta machine 10 MV. The neutron source strengths Q are calculated. It was found to be 0.2×1012 n Gy−1 X-ray at the isocenter. This work represents simple, low cost, and accurate methods of measuring fast neutrons dose and energies.

Neutron beam preparation with Am–Be source for analysis of biological samples with PGNAA method

Material analysis with prompt gamma neutron activation analysis (PGNAA) requires a proper geometrical arrangement for equipments in laboratory. Application of PGNAA in analysis of biological samples, due to small size of sample, needs attention to the dimension of neutron beam. In our work, neutron source has been made of 241Am–Be type. Activity of 241Am was 20 Ci which lead to neutron source strength of 4.4 × 107 neutrons per second. Water has been considered as the basic shielding material for the neutron source. The effect of various concentration of boric acid in the reduction of intensity of fast and thermal components of the neutron beam and gamma ray has been investigated. Gamma ray is produced by (α, n) reaction in Am–Be source (4.483 MeV), neutron capture by hydrogen (2.224 MeV), and neutron capture by boron (0.483 MeV). Various types of neutron and gamma ray dosimeters have been employed including BF3 and NE-213 detectors to detect fast and thermal neutrons. BGO scintillation detector has been used for gamma ray spectroscopy. It is shown that the gamma and neutron radiation dose due to direct beam is of the same magnitude as the dose due to radiation scattered in the laboratory ambient. It is concluded that 14 kg boric acid dissolved in 1,000 kg water is the optimum solution to surround the neutron source. The experimental results have been compared with Monte Carlo simulation.

SU‐E‐T‐343: Monte Carlo Study on a Novel Direction‐Selective Flattening Filter System

Purpose: Monte Carlo evaluation of a newly‐designed narrow steel cone flattening filter, which functions via the direction‐selective attenuation of the photons produced in the target. Methods: Using BEAMnrc and DOSXYZnrc (EGSnrc), novel 6 and 15 MV bremsstrahlung beams were modeled for a Siemens Primus linac. Under flattening filter free (FFF) conditions, the classical flattening filter was omitted. For the direction‐selective filter (DSF) configuration, an axis‐centered bullet‐shaped narrow steel cone filter is now mounted directly below the graphite electron absorber of the tungsten target. The neutron source strength was modeled by assessing photoneutron production at all tungsten surfaces hit by photons, assuming total photon absorption and using the macroscopic photoneutron cross section of tungsten. Results: Photon spectra for FFF and DSF conditions are less filtered compared to classical flattening filter conditions, while depth dose curves are very similar. Under DSF conditions, field flatness has been achieved for a circle of diameter 15 cm at 100 cm SSD, and the normalized peripheral dose was reduced by factor 1.5. For FFF and DSF, the dose rate at 10 cm water phantom depth was enhanced by factors 2 and 1.7 (6 MV) and 4 and 2.5 (15 MV) respectively, while the corresponding beam head photon leakage dose remained unchanged. The photoneutron source strength produced/Gy was reduced by factors of 3.5 (FFF) and 2.8 (DSF). Conclusions: With the DSF system, the simplicity of treatment techniques and workflow is maintained, because field flatness up to 15 cm field diameter is achieved at all depths, while the advantages characteristic for FFF conditions, the dose‐rate enhancement, the reduction of photon and neutron leakage from the beam head/Gy in the treatment volume, and the reduction of the normalized dose levels in the far periphery are maintained.

Effect of thermal neutrons on fusion power measurement using the microfission chamber in ITER

A microfission chamber (MFC) provides time-resolved measurements of global neutron source strength and fusion power in ITER. The MFC is a pencil-sized gas counter containing the fissile material, 235U and will be installed behind blanket modules at upper and lower outboard positions. Measurements of the neutron source strength could be affected by cooling water in branch pipes installed near the MFC. The effect of the branch pipes upon the MFC is assessed through neutron transport calculation using MCNP 5. Results indicate a significant increase in the MFC sensitivity (up to approximately 40% higher) due to the branch pipe. The increase in the MFC response is caused by the slowing down of the neutrons due to the cooling water in the branch pipes. The effect of thermal neutrons on MFC response is especially significant. One possible solution to reduce the effect is to cover the MFC with a material that absorbs thermal neutrons such as cadmium. The ways in which the absorbent material may affect MFC response is analyzed through neutron transport calculation. Results indicate that the increase in the MFC response can be reduced to < 10% through cadmium coating.

Development of in-vessel components of the microfission chamber for ITER

Microfission chambers (MFCs) will measure the total neutron source strength in ITER. The MFCs will be installed behind blanket modules in the vacuum vessel (VV). Triaxial mineral insulated (MI) cables will carry signals from the MFCs. The joint connecting triaxial MI cables in the VV must be considered because the MFCs and the MI cables will be installed separately at different times. Vacuum tight triaxial connector of the MI cable has been designed and a prototype has been constructed. Performance tests indicate that the connector can be applied to the ITER environment. A small bending-radius test of the MI cable indicates no observed damage at a curvature radius of 100 mm.

Secondary Radiation Dose during High-energy TBI

Secondary Radiation Dose during High-energy TBI