Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor

[EN] All materials exposed to some neutron flux can be activated independently of the kind of the neutron source. In this study, a nuclear reactor has been considered as neutron source. In particular, the activation of control rods in a BWR is studied to obtain the doses produced around the storage pool for irradiated fuel of the plant when control rods are withdrawn from the reactor and installed into this pool. It is very important to calculate these doses because they can affect to plant workers in the area. The MCNP code based on the Monte Carlo method has been applied to simulate activation reactions produced in the control rods inserted into the reactor. Obtained activities are introduced as input into another MC model to estimate doses produced by them. The comparison of simulation results with experimental measurements allows the validation of developed models. The developed MC models have been also applied to simulate the activation of other materials, such as components of a stainless steel sample introduced into a training reactors. These models, once validated, can be applied to other situations and materials where a neutron flux can be found, not only nuclear reactors. For instance, activation analysis with an Am-Be source, neutrography techniques in both medical applications and non-destructive analysis of materials, civil engineering applications using a Troxler, analysis of materials in decommissioning of nuclear power plants, etc.Ródenas Diago, J. (2017). Application of the Monte Carlo method to estimate doses due to neutron activation of different materials in a nuclear reactor. Radiation Physics and Chemistry. 140:442-446. doi:10.1016/j.radphyschem.2017.02.015S44244614

Analysis of nuclear reactions to determine the radionuclides generated and its activity in various devices

[EN] Nuclear reactions can generate radionuclides and it is necessary to know the activity produced in the reaction. Most of the activation reactions are induced by neutrons since the addition of one neutron produces normally an instability in the nucleus, but other particles like protons can also be used to produce radionuclides. Nuclear reactors and accelerators are some of the most used devices to host nuclear reactions to produce radionuclides for various applications. The MCNP5 code based on the Monte Carlo (MC) method can be used to estimate the activity generated in the reaction. A summary of the simulation of neutron reactions is presented in the paper. The comparison of simulation results with experimental measurements allows the validation of developed models.Ródenas Diago, J.; Verdú Martín, GJ. (2020). Analysis of nuclear reactions to determine the radionuclides generated and its activity in various devices. Radiation Physics and Chemistry. 167:1-6. https://doi.org/10.1016/j.radphyschem.2019.05.011S16167Ródenas, J., Gallardo, S., Abarca, A., & Juan, V. (2010). Estimation of the activity generated by neutron activation in control rods of a BWR. Applied Radiation and Isotopes, 68(4-5), 905-908. doi:10.1016/j.apradiso.2009.09.059Ródenas, J., Abarca, A., Gallardo, S., & Sollet, E. (2010). Validation of the Monte Carlo model developed to assess the activity generated in control rods of a BWR. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 619(1-3), 258-261. doi:10.1016/j.nima.2009.10.084Ródenas, J., Gallardo, S., Abarca, A., & Juan, V. (2010). Analysis of the dose rate produced by control rods discharged from a BWR into the irradiated fuel pool. Applied Radiation and Isotopes, 68(4-5), 909-912. doi:10.1016/j.apradiso.2009.09.060Ródenas, J., Gallardo, S., Weirich, F., & Hansen, W. (2014). Application of dosimetry measurements to analyze the neutron activation of a stainless steel sample in a training nuclear reactor. Radiation Physics and Chemistry, 104, 368-371. doi:10.1016/j.radphyschem.2014.05.01

Determinación de los parámetros de un reactor nuclear necesarios para obtener radioisótopos con actividad exenta

[ES] El uso de radionucleidos como marcadores en un proceso industrial es muy útil. El producto marcado debe contener una actividad por debajo de los valores exentos para que pueda considerarse después del proceso como un residuo convencional. Además, la dosis potencialmente recibida por los trabajadores durante el proceso tiene que estar por debajo de los límites, ya que no trabajan en zona controlada. El objetivo de este trabajo es analizar la influencia de varios parámetros en una activación neutrónica. El código MCNP, basado en el método Monte Carlo, se ha utilizado para simular la activación de diferentes materiales en un reactor nuclear de investigación. La producción de Na-24, Mo-99 y Sb-124 mediante reacciones (n, ¿) se ha simulado evaluando los parámetros que influyen en la actividad generada que deberá ser exenta según la legislación vigente. Los resultados permiten verificar que es posible utilizar la activación neutrónica para obtener Na-24, Mo-99 y Sb-124 con actividades exentas, siempre que se ajusten los parámetros analizados.Ródenas Diago, J.; Aledó Cuenca, S. (2019). Determinación de los parámetros de un reactor nuclear necesarios para obtener radioisótopos con actividad exenta. Sociedad Nuclear Española. 1-8. http://hdl.handle.net/10251/180967S1

Análisis mediante el método de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactor BWR

[ES] Las barras de control de un reactor de agua en ebullición (BWR) están sometidas a un flujo neutrónico y por tanto, resultan activadas durante su permanencia en el núcleo del reactor. La activación se produce especialmente en los componentes del acero inoxidable y en las impurezas. La actividad generada da lugar a una dosis alrededor de la barra, sin importancia mientras está en el reactor, pero que debe tenerse en cuenta cuando se extrae del mismo. Las barras extraídas se almacenan en colgadores situados en las piscinas de almacenamiento del combustible irradiado de la central. Cada colgador aloja 12 barras de control y se disponen de modo que haya al menos tres metros de agua por encima de los cabezales de las barras de control. La dosis potencialmente recibida por los trabajadores profesionalmente expuestos que se encuentren en las inmediaciones de la piscina de almacenamiento debe calcularse para asegurar la adecuada protección de los mismos. Esta dosis puede disminuirse de modo importante si se cambia la disposición de las barras en los colgadores.Ródenas Diago, J.; Gallardo Bermell, S. (2011). Análisis mediante el método de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactor BWR. Revista de Física Médica. 12(Supl):508-508. http://hdl.handle.net/10251/99443S50850812Sup

Unfolding X-ray spectra using a flat panel detector. Determination of the accuracy of the method with the Monte Carlo method

[EN] The primary X-ray spectrum depends on different parameters such as high voltage, filament current, high voltage ripple, anode angle and thickness of filter material. The objective of this work is to determine whether the unfolding technique based on the Tikhonov regularization method is accurate enough to estimate the X-ray spectrum when slight changes in the operation variables are considered. In this frame, several X-ray spectra are considered (extracted from the IPEM78 Catalogue Report) varying the main operation variables of the X-ray tube (high voltage, voltage ripple, filter thickness and filter material). With those spectra, the corresponding absorbed dose curves are obtained by simulation with a MCNP5 model reproducing a flat panel detector and a PMMA wedge. Once the absorbed dose curves are simulated and applying the unfolding Tikhonov regularization method, the unfolded spectrum is obtained, which is finally compared with the theoretical one (IPEM78 Catalogue Report). Discrepancies between unfolded and primary X-ray spectra can be attributed to the fact that this is an ill-posed problem, and the unfolding of the spectrum is strongly affected by the method used to improve the conditioning of the response function (response matrix).Gallardo Bermell, S.; Ródenas Diago, J.; Verdú Martín, GJ. (2019). Unfolding X-ray spectra using a flat panel detector. Determination of the accuracy of the method with the Monte Carlo method. Radiation Physics and Chemistry. 155:233-238. https://doi.org/10.1016/j.radphyschem.2018.09.014S23323815

Application of dosimetry measurements to analyze the neutron activation of a stainless steel sample in a training nuclear reactor

All materials present in the core of a nuclear reactor are activated by neutron irradiation. The activity so generated produces a dose around the material. This dose is a potential risk for workers in the surrounding area when materials are withdrawn from the reactor. Therefore, it is necessary to assess the activity generated and the dose produced. In previous works, neutron activation of control rods and doses around the storage pool where they are placed have been calculated for a Boiling Water Reactor using the MCNP5 code based on the Monte Carlo method. Most of the activation is produced indeed in stainless steel components of the nuclear reactor core not only control rods. In this work, a stainless steel sample is irradiated in the Training Reactor AKR-2 of the Technical University Dresden. Dose measurements around the sample have been performed for different times after the irradiation. Experimental dosimetric values are compared with results of Monte Carlo simulation of the irradiation. Comparison shows a good agreement. Hence, the activation Monte Carlo model can be considered as validated.Ródenas Diago, J.; Gallardo Bermell, S.; Weirich, F.; Hansen, W. (2014). Application of dosimetry measurements to analyze the neutron activation of a stainless steel sample in a training nuclear reactor. Radiation Physics and Chemistry. 104:368-371. doi:10.1016/j.radphyschem.2014.05.013S36837110

Estudio de la activación neutrónica del acero inoxidable en un reactor nuclear

Durante la operación de un reactor nuclear, diferentes componentes pueden activarse por reacciones neutrónicas. La actividad así generada produce una dosis que es un riesgo potencial para los trabajadores del entorno. Se ha simulado mediante los códigos MCNP y CINDER’90 dicha activación en una pieza de acero y comparado los valores obtenidos con mediciones experimentales. Se verifica la equivalencia de ambos métodos para los cálculos de activación neutrónica y evolución de la tasa de dosis con el tiempo de enfriamiento.Lázaro Roche, I.; Ródenas Diago, J.; Marques, J. (2013). Estudio de la activación neutrónica del acero inoxidable en un reactor nuclear. Sociedad Nuclear Española. http://hdl.handle.net/10251/71624

Using lattice tools and unfolding methods for hpge detector efficiency simulation with the Monte Carlo code MCNP5

In environmental radioactivity measurements, High Purity Germanium (HPGe) detectors are commonly used due to their excellent resolution. Efficiency calibration of detectors is essential to determine activity of radionuclides. The Monte Carlo method has been proved to be a powerful tool to complement efficiency calculations. In aged detectors, efficiency is partially deteriorated due to the dead layer increasing and consequently, the active volume decreasing. The characterization of the radiation transport in the dead layer is essential for a realistic HPGe simulation. In this work, the MCNP5 code is used to calculate the detector efficiency. The F4MESH tally is used to determine the photon and electron fluence in the dead layer and the active volume. The energy deposited in the Ge has been analyzed using the *F8 tally. The F8 tally is used to obtain spectra and to calculate the detector efficiency. When the photon fluence and the energy deposition in the crystal are known, some unfolding methods can be used to estimate the activity of a given source. In this way, the efficiency is obtained and serves to verify the value obtained by other methods.Querol Vives, A.; Gallardo Bermell, S.; Ródenas Diago, J.; Verdú Martín, GJ. (2015). Using lattice tools and unfolding methods for hpge detector efficiency simulation with the Monte Carlo code MCNP5. Radiation Physics and Chemistry. 116:219-225. doi:10.1016/j.radphyschem.2015.01.027S21922511

Dosimetric characterization of a brachytherapy applicator using MCNP5 modelisation and in-phantom measurements

[EN] A gynaecological applicator consisting of a metallic intra-uterine tube with a plastic vaginal applicator and an HDR Ir-192 source have been simulated with MCNP5 (Monte Carlo code). A solid phantom has been designed to perform measurements around the applicator with radiochromic films. The isodose curves obtained are compared with curves calculated with the F4MESH tally of MCNP5 with a good agreement. A pinpoint ionization chamber has been used to evaluate dose at some reference points. (c) 2009 Elsevier Ltd. All rights reserved.Gerardy, I.; Ródenas Diago, J.; Van Dycke, M.; Gallardo Bermell, S.; Ceccolini, E. (2010). Dosimetric characterization of a brachytherapy applicator using MCNP5 modelisation and in-phantom measurements. Applied Radiation and Isotopes. 68(4):735-737. doi:10.1016/j.apradiso.2009.10.018S73573768

Estudio de la reconstrucción del espectro primario de rayos X a partir de la simulación de detectores de semiconductor

[ES] El presente trabajo pretende estudiar la idoneidad de tres detectores de semiconductor: Germanio, Silicio y Teluro de Cadmio para reconstruir el espectro primario de los equipos de radiodiagnóstico en distintos intervalos de energía. Para ello, se utiliza un método mixto experimental-Monte Carlo (MC) y un método de reconstrucción. Se ha utilizado el programa MCNP5 con el fin de simular todo el proceso físico que tiene lugar entre la fuente de rayos X, el espectrómetro y cada uno de los detectores. Tras un estudio previo de los métodos matemáticos de reconstrucción: el Modificado Truncado de Descomposición en Valores Singulares (MTSVD), el Amortiguado de Descomposición en Valores Singulares (DSVD) y el de Tikhonov, se ha escogido para el presente trabajo este último por ser el que mejor aproxima el espectro reconstruido al espectro teórico.Querol, A.; Gallardo Bermell, S.; Ródenas Diago, J.; Verdú Martín, GJ. (2011). Estudio de la reconstrucción del espectro primario de rayos X a partir de la simulación de detectores de semiconductor. Revista de Física Médica. 12(S):377-378. http://hdl.handle.net/10251/99473S37737812