In this study, a Monte Carlo model has been developed for a Cerenkov-based fiber-optic gamma-ray sensor (CFOGRS) using the GEANT4 simulation toolkit. The detection material for gamma rays in CFOGRS is the transparent silica core of the optical fiber, which is also used for optical signal propagation. The model implemented with the GEANT4 includes the transport process of gamma rays, as well as the physical processes of Compton scattering, the Cerenkov effect, and optical photon propagation within the optical fiber. The model also simulated the applicability of the CFOGRS in a radiation environment by using the Monte Carlo code of GEANT4.

The efficient and accurate burn-up measurement of the spherical fuel element is the key component of the operation of the pebble bed high temperature gas-cooled reactor. The accuracy of the method that determine burnup by the activity of Cs-137 degrades due to operation characteristics of HTR-10. HTR-10, as an test reactor, operated on and off during the past years. It stayed shutdown more than power operation. In order to improve the measurement accuracy of Cs-137 activity and enhance the possibility to detect radionuclides with low activity, which can be used to correct the classic burnup assay method, a new measurement system is now discussed using anti-coincidence technology, which suppresses the Compton plateau. In this paper, Geant4 is used to simulate the anticoincidence measurement process taking high purity germanium γ-ray spectrometer as main detector and plastic scintillator as the annular detector. By analyzing the signal to noise ratio in different detection scenarios with all kinds of shape parameters of the annular detector, the annular detector with the best anti-coincidence effect are optimaized. The above research results provide an important theoretical basis for the construction of online burn-up measurement system based on anti-Compton technology.

This Paper presents an improved estimation of reactor core baffle temperature distribution, during operation, at the nominal power level to address swelling problems of the reactor internals. Swelling is the main limiting factor in the reactor core internals long term operation of VVER-1000 nuclear units. The material irradiation-induced swelling and creep models are very sensitive to temperature distribution in metal, thus a more detailed analysis of the core baffle metal thermohydraulic cooling characteristics is required. A framework for CFD analysis of VVER-1000 reactor baffle cooling is presented. Firstly, an analytical model was developed to obtain boundary conditions and simplify CFD analysis, i.e. the real geometry of the cooling channels was replaced by equivalent elements, the core was presented as porous body with the appropriate characteristics. Secondly, the CFD analysis was performed using 60–degree symmetry, which included: core, baffle and core barrel, it is limited by the height of the baffle. Core is simplified as a homogeneous body with considering of spatial volumetric energy release. Core baffle is presented as monolithic body with considering of gamma-ray heat generation. Model includes a simplified geometry of connecting studs, considering cooling flow of the coolant through the nuts grooves. Calculated convection coefficient and temperature are in good agreement with analytical model, and give a more convenient result comparing to RELAP5/mod3.2. Obtained results were used to estimate baffle swelling process. Due to the less conservative results in temperature distribution swelling and creep deformations significantly decreased.

Excessive gamma rays will be emitted when a nuclear power plant is under the refueling overhaul, leading to a certain number of hotspots. To meet the shielding requirements of these hotspots, a new type of flexible shielding material with a “sandwich structure” was developed. This kind of material has been adopted in a mega-kilowatt pressurized water reactor (PWR) nuclear power plant as the first demonstration project. Based on the characteristics of the new flexible shielding material, the process design program of the product application was founded, and the new material was compared with the traditional lead shielding material. At last, the installation and removal time together with the shielding effect were analyzed according to the first demonstration application.

For nuclear fuel reprocessing plants, some materials with high activity have been collected in the product container in the process of spent fuel reprocessing. In order to ensure the staff’s occupational radiation safety, an analysis of product container radiation shielding and source terms is crucial to determine the total activity limit and the activity limit range of different rays while reprocessing radioactive materials in the product container, so as to determine the working time limitation. A basic calculation result shows that generally alpha particles are not taken into consideration when calculating its occupational radiation effect for product container. In order to reduce the radiation exposure caused by neutron and gamma ray, radiation nuclides such as actinides and activated products which have a high neutron and gamma yield, and also the total nuclide amount should be controlled. In addition, the purification of nuclide emitting pure beta ray such as 90 Sr, 90 Y, 106 Ru, 106 Rh should be strictly supervised in order to restrict the beta ray exposure and its bremsstrahlung effect.

Current gamma-ray spectrum analysis method uses a preset system response matrix to improve the resolution of gamma-ray spectrum. However, the system response matrix may not be available or biased due to limitation of experiment conditions, which can degrade the accuracy of gamma-ray spectrum analysis. To solve the problem, a new reconstruction method based on blind deconvolution and sparsity constraint is proposed to improve the resolution of gamma-ray spectrum in this study. The proposed method models the modulation of spectrometer as a convolution operation and reconstructs the high resolution spectrum as well as the convolution kernel simultaneously. Lp-norm based sparsity constraint is imposed to stabilize the demodulation of spectrometer and reduce the background oscillations, so that the resolution can be enhanced. The results of both numerical simulation and experiments demonstrate that the proposed method can effectively improve the resolution of gamma-ray spectrum and reduce background oscillations without any aid of system response matrix.

TlBr is a kind of semiconductor material. Due to its promising physical properties and can be used at room temperature, it is continually studied as X and gamma ray detectors candidate material. Both of its atomic number and density are high. It also has large band-gap (B = 2.68eV) and low ionization energy. TlBr device exhibits high detection efficiency and excellent energy resolution. It can be easily fabricated or compacted in small housing. So it is a reasonable selection in the fields of nuclear material inspection and safeguards property, national security, spatial and high energy physics researches. The paper investigates the TlBr radioactive detector development and fabrication procedures. The processing detail information and signals collection are emphasized in different section. The prototype detectors were irradiated by Am-241 and corresponding spectrum was obtained. The photoelectric peak at 59.5keV is distinguished visible and the best resolution at 59.5keV is 4.15keV (7%).

Boron Neutron Capture Therapy (BNCT) is a kind of the targeted therapy with two element. It can kill the cancer cells while the effect on normal cells is very small, and it is suitable for the treatment of the various stage cancer so it will be the ideal radiotherapy for cancer treatment in the future. And Commercial Miniature Neutron Source Reactor (C-MNSR) was designed and constructed by CIAE, which is used for Neutron Activation Analysis (NAA), Training and teaching. The reactor with thermal power 27kW is an under-moderated reactor with pool-tank type, U-AL alloy with High Enriched Uranium (HEU) as fuel, light water as coolant and moderator, and metal beryllium as reflector. The fission heat produced by the reactor is removed by the natural circulation. Design C-MNSR with a epi-thermal neutron beam for BNCT is studied while the conversion from HEU to LEU (Low Enrichment Uranium) ( 235 U percent≤20%) is carried on. As it has the advantages of MNSR safety, economy, easy operation and its application, and it can improve the epi-thermal neutron flux density and meet the requirements of BNCT. The fuel cage of C-MNSR with size of φ230×248mm in the reactor core, there are ton rows of 355lattices are concentrically arranged, the central lattice is reserved for central control rod, and four tie rods are uniformly arranged at the eighth row which link the upper and lower grid plates, the rest 350 fuel lattices are for fuel pins or dummies. The diameter of the fuel meat is 4.3mm, the height is 230mm, with Uranium enrichment is 17%; the diameter of the fuel element is 5.5mm, the height is 248mm. The frame design of the epithermal neutron beam is: Fluental material used as neutron moderation layer with its thickness is 50cm and its density is 2.85g/cm3; Cd with thickness of 0.1cm used as thermal neutron absorption layer, Lead with thickness of 10cm used as gamma ray shielding layer. And the neutron collimator parts is a composition of graphite, Cd and polythene with boron. The total length of the beam is 114.5cm, and the distance from the exit of the beam to the core is 130cm. The results show that the epithermal neutron flux density at the exit is 1.58 × 10 9 n·cm-2·s-1 at full power of 27kW. and the fast neutron density at the exit is 5.45 × 10 7 n · cm-2 · s-1 at full power. Fast neutron dose contamination (Df/ φepi) is 2.88 × 10 −11 Gy · cm 2 · n −1 and gamma dose contamination (Dγ/φepi) 2.18× 10 −14 Gy·cm 2 ·n −1 .

The bipolar operational amplifier is an important kind of electronic device used in the nuclear reactor control system and its radiation damage threatens the reliability of nuclear operation. In order to explore the degradation of the bipolar operational amplifier induced by reactor neutron and gamma irradiation, radiation experiments were accomplished on particular bipolar operational amplifiers OP07 manufactured by Texas Instruments and Analog Devices . The amplifiers were irradiated in three different environments respectively: individual neutrons, individual gamma rays and mixed irradiation of reactor neutrons and gamma rays simultaneously. The results indicate that for both types of bipolar operational amplifiers, the degradations of the input bias current and the input offset voltage induced by the mixed irradiation of reactor neutrons and gamma rays simultaneously are notably more severe than the simple sum of neutron and gamma degradations measured individually. The mixed irradiation of reactor neutron and gamma rays simultaneously leads to degradation enhancement compared to individual ones and induce ionizing/displacement synergistic effects in the amplifiers. Enough attention should be paid to the degradation enhancement induced by the mixed irradiation for devices may exhibit unexpected failure during reactor operation and accelerate the aging of the nuclear reactor control system.

Mixed neutron and gamma radiations require different shielding materials as their interaction with materials is different. Composites were developed in order to combine the shielding capabilities of different materials. However, their homogeneity is difficult to be assured which can lead to pinholes where radiation can penetrate. To avoid this problem, several materials arranged in layers can be used to shield against mixed radiations. Since the multilayer shielding can be made from any material in many configurations, the ant colony optimization (ACO) is a promising method because it deals with combinatorial optimization problems. The candidate materials are HDPE, boron, cadmium, gadolinium, tungsten, bismuth, and iron. Preliminary MCNP simulations were done to observe the effect of arrangements, thicknesses, and types of materials on the radiation spectrum. It was found that: (1) the final layer should be made of high density material, (2) an increase beyond certain thicknesses did not result in a significant increase in attenuation, and (3) there should be an optimum combination of material that can effectively shield against both neutrons and gamma rays.

Alpha radiation emitting radon (Rn) gas seepage into homes in the USA leads to over 21,000 annual lung cancer deaths (according to the US-Environmental Protection Agency, EPA) leading to mandatory monitoring for Rn throughout the USA. In the nuclear industry alpha emitting radionuclides in air (e.g., in spent fuel reprocessing) also constitute a major safety and security-safeguards related issues. Purdue University, along with Sagamore Adams Laboratories LLC, is developing the tensioned metastable fluid detector (TMFD) technology for general-purpose alpha-neutron-fission spectroscopy. This paper focuses on rapid, high-efficiency detection of Rn and progeny in air using the novel TMFD technology; Rn and progeny isotopes in air are sparged through the TMFD detection fluid (to entrap the radioactive gas), which is then placed under a metastable state. Through tailoring the metastable fluid state, an audible and visible cavitation detection event is created and readily detected from transient bubble formation. Changing the tensioned state allows for the spectroscopic differentiability of Rn and its daughters which can be used to actively measure the equilibrium between the parent and daughter products. Such a technique can also be used to monitor the atmosphere in critical nuclear facilities for contamination from other alpha emitting isotopes (e.g., Pu, Cm, U...). TMFDs offer the unique ability for high intrinsic efficiency (>95%) alpha-neutron-fission fragment detection, while remaining blind to background beta and gamma radiation (qualified to >3.8×10 8 Bq m −3 using a dissolved 32 P beta source, and also via gammas from a 53 R/hr 137 Cs gamma source). Immunity to beta and gamma is beneficial for the discrimination of buildup of beta-emitting Thoron and Rn progeny in the detection fluid allowing for reusability. This paper will discuss the research results pertaining to detection of Radon and progeny in air, for concentrations between 74 Bq m −3 (2 pCi/L) and 740 Bq m −3 (20 pCi/L). The system measures a radon concentration between these levels to within ±15% intrinsic relative error (IRE) within 24 hours meeting the standards outlined by the American Association of Radon Scientists and Technicians-National Radon Proficiency Program (AARST-NRPP) Device Evaluation Program. Precision evaluation was also performed and the relative standard deviation defined by the AARST-NRPP was <5% exceeded the requirement of 25%. Ambient temperature effects were assessed at 10 °C and at 27 °C, which revealed a large increase in collection efficiency with decreasing sampling temperature and slight increase with increasing sampling temperature. Temperature effects on sensitivity thresholds and volumetric expansion were measured and used to compensate for variability in temperature over time. Blind testing with the help of Bowser-Morner Radon Reference Laboratory was performed and succeeded in accurately determining the Rn in air concentration to within 20% within only 6h of sampling. Finally, a 48-hour based collection time has also been developed for use in dwellings where Rn in air concentrations may vary in a day. Overall, the reproducibility and precision of TMFD technology is found to allow for an efficient, cost-effective, reliable, and environmentally friendly means of Rn and progeny detection, and by extension for use for general actinide in air monitoring for the nuclear industry.

Hydration is a main process following mixing of cement with water, producing a chemical reaction between cement particles and water molecules known as Calcium-Silica-Hydroxide (C-S-H). The C-S-H acts as a binder between coarse and fine mix aggregates. During mixing, capillary voids are formed where water can easily become trapped. The van der Waals force being the principal source of strength between such compounds, voids become detrimental to long-term strength of concrete because they decrease the amount of surface area aggregate and cement particles have to come into contact with each other. An overall decrease in both pores size and their distribution in concrete increase density, therefore increasing the van der Waals force between these particles. In order to reduce capillary pores size and their distribution, in this paper we present a result from preliminary investigation on how gamma ray exposure during the early stages of the curing process affect the concrete mixture strength values. The gamma ray enhanced curing is expected to break the water molecules trapped in capillary pores into H + and OH − , allowing these ions to escape and therefore result in densifying the concrete mixture. Concrete cubes of 5cm 3 in volumes are exposed to isotropic 137 Cs gamma emitting source. The absorbed dose in cubes is measured as well as calculated using MCNP6. Each concrete cube is exposed to a 630 MBq 137 Cs source for seven continuous days. The absorbed dose is experimentally measured using a Landauer nanoDot system. The average measured dose is 1.12 Gy at the front and 0.33 at the back of a concrete, totaling 0.79 Gy. The average measured dose is 0.98 Gy at the front and 0.29 Gy at the back, totaling 0.69 Gy of absorbed dose, thus showing a good agreement between the numerical predictions and experimental measurements. Compressive strength of concrete mixes exposed to gamma rays during the curing time and those cured without exposure to gamma radiation are tested. Preliminary data shows that the average compressive strength for gamma ray enhanced curing of concrete mixes is in order of 9,000 psi while the average com-pressive strength for the cubes cured in dry air is on average about 5,000 psi. A number of new experiments is planed until the time of the Conference in providing more exact answer to the question: does gamma ray enhanced curing of concrete increase its strength and reduces the curing time?

In line with the cooperative project “Non-invasive Condition Monitoring of Nuclear Reactors for the Detection of Level Change and Deformation of the Core” between the Technical University Dresden and the Institute of Process Technology, Process Automation and Measuring Technology (IPM) of Zittau/Goerlitz University of Applied Sciences, a measuring system for the core state diagnosis during a core melt accident in the reactor pressure vessel of a pressurized water reactor is going to be developed. The operational principle of this system is based on the non-invasive measurement of continuously changing gamma radiations (caused by the shifting of melted materials and fission products) outside of the reactor pressure vessel by means of several gamma radiation sensors. The sensors are arranged over the height of the core and the lower head. By using computer based and real-time capable methods for evaluation of the measured gamma radiations conclusions about the core state can subsequently be drawn. This paper includes a description of a concept as well as several methods for the core state diagnosis during a core melt accident. The main part is the analysis of the methods for the core state diagnosis based on different evaluation criteria.

The gamma radiation emitted by a nuclear reactor core might contain information about the reactor state. This information may be used in a monitoring system for severe accidents. The Technische Universität Dresden and the Zittau/Görlitz University of Applied Sciences are currently carrying out feasibility studies for the development of such a system in a collaborative effort. As one part of such feasibility studies we performed Monte Carlo simulations on a simplified model of a generic pressurized water reactor. For a set of states which represent scenarios of a coolant level decrease and core melt, the gamma radiation distribution outside the reactor pressure vessel has been computed. The results are presented in this paper. They indicate that different coolant levels yield different gamma radiation distributions, and that an accumulation of corium inside the lower head is detectable from the outside.

A high-radiation resistant optical fiber has been developed in order to investigate the interiors of the reactor pressure vessels and the primary containment vessels of the Fukushima Daiichi Nuclear Power Station. The radiation resistance of an optical fiber was improved by increasing the amount of hydroxyl up to 1000 ppm in pure silica fiber. We have tried to apply the optical fiber for remote imaging technique by means of fiberscope. The improved image fiber consists of common cladding and a large number of fiber cores made from pure silica that contains 1000 ppm hydroxyl. The transmissive rate of an infrared image was not affected after the irradiation of 1 MGy. The radiation resistant optical fiber is available for remote ultimate analysis by laser induced breakdown spectroscopy (LIBS) in order to identify whether a material is fuel debris or not. We have developed the fiber-coupled LIBS system to detect plasma emission efficiently in near-infrared region. In addition, we have performed a gamma ray dose rate measurement using an optical fiber of which scintillator is attached to the tip. As a result, the concept of applicability of a probing system using the high-radiation resistant optical fibers has been confirmed.

The environmental radiation level and radioactivity were analyzed in the Ulsan which is close to Nuclear Power Plants (NPPs). Ulsan is a Metropolitan city with a population of about 1,100,000 residents, where the density of NPPs is the highest in Korea. Environmental radiation monitoring in Ulsan has been performed since the radiation monitoring facility was opened in August 2012 and the radioactivity was measured and analyzed in the environmental samples for five months in Ulsan. The measurement of environmental radiation and radioactivity was carried out periodically for various samples including airborne dust, fallout, tap water, and precipitation. Also, sampled were the soil, rice, cabbage, and indicator plants such as Artemisia and pine needles. The basic data from the radioactivity analyses brought a fundamental data for assessing the internal exposure dose for the public due to the intake of the radio nuclides included in those samples. On the other hand, the gamma radiation in the environment was measured continuously by using an environmental radiation monitor of the ion chamber type. Actually, the measurement of gamma radiation is thought to provide the base-line data on environmental radiation/radioactivity for radiological emergencies and the real-time information compared with the background radiation. The measurement analyses showed that the radiation level of the Ulsan area is kept at a normal background level in spite of the operation of many nuclear power plants near it.

The gamma-ray dose calculation is essential for the radiation shielding of pressurized water reactor (PWR) spent fuels. Homogenization modeling of fuel pin lattices for typical PWR spent fuel pins is regularly applied on the radiation protection calculation of gamma-ray dose in an air medium. However, depending on the size of the homogenized lattice and the location of the detectors, under-estimation or over-estimation of the gamma-ray dose due to the homogenization modeling can be obtained with respect to the detailed heterogeneous model. In previous published results from MCNP-4A and 4C calculations on gamma-ray dose from spent PWR fuel pins, very different homogeneous to heterogeneous (Hom/Het) ratios were reported. In this study these Hom/Het ratios have been re-evaluated and benchmarked by using the TRIPOLI-4 Monte Carlo transport code. The new TRIPOLI-4 mesh tally capabilities have also been applied to calculate the radial and axial gamma-ray dose distribution. With the recently upgraded TRIPOLI-4 display tool, the dose rate maps and the isodose rate curves around a spent PWR fuel assembly have been established.

A low-background gamma spectrometer consists of a high-performance gamma detector and a low-background chamber. It is widely used to monitor the radiation level of the environment and to identify the species of the radiological source. It is especially important for the analysis of the nuclear accident. Usually a high purity Germanium detector (HPGe) is used as a gamma ray detector. In order to enhance the detecting accuracy and sensitivity, it is essential to improve the performance of the gamma detector. In recent years, a clover detector composed of four coaxial HPGe crystals appear and is widely utilized in nuclear physics experimental research. Because of the larger dimensions and segmented structure, it displays outstanding characteristics different from traditional HPGe detectors. With a clover detector as the main detector and the HPLBS1 chamber of ORTEC as the lead chamber, the low-background gamma spectrometer is simulated by the Monte Carlo toolkit GEANT4, where the interaction processes of gamma ray provided by the GEANT4 physics list is used. The detecting performance of the low-background gamma spectrometer such as detecting efficiency and peak-total ratio are given. The results indicate that low-background gamma spectrometer with a clover as the main detector has better characteristic than that with HPGe as a main detector traditionally.

The coefficient of detector efficiency is an essential while we use lung counting system to assess the internal exposure dose. This paper used an anthropomorphic phantom which was made accordance with Chinese Reference Man to calibrate our system. We cooperated with China Institute for Radiation Protection to manufacture the phantom which was made of tissue equivalent materials. In the lung there are 246 club-shaped uranium sources. Each source includes (20.12±0.32) Bq 238 U and (0.53±0.01) Bq 235 U. There are 4 parts in this paper. We start with a brief introduction to our lung counting system. Then the method of how to manufacture the lung and the result of the parameters such as density, linear absorption coefficient of the lung are provided in detail. The efficiency curve under different Chest Wall Thickness (CWT) is obtained. In our application, about 22 workers have been measured finally. We find in our research that the Minimum Detectable Activity (MDA) of natural uranium changes from 2.5mg to 5.9mg as the CWT varies from 16.2mm to 40.5mm for our lung counting system in 4000s detecting time by analyzing 63.3keV photo-electricity peak of gamma ray. Based on the least limited value of annual inhalation is about 220mg while the lung slowly absorbed uranium aerosol with particles about 1 μm, we may conclude that our lung counting system can be well used to detect the activity of natural uranium in a short time after a worker inhaled uranium aerosol.

A new scintillator, CLYC, has been investigated for possible use in neutron spectrometry. This sensor provides neutron detection for both thermal and fast neutrons from the reactions 6 Li(n,α) and 35 Cl(n,p), respectively. This work primarily focuses on the detection of fast neutrons since there is currently no sensor that can accurately and efficiently provide information about their incident neutron energy. Conventional methods of fast neutron detection have been based on utilizing materials that use the elastic scattering process of neutrons on 1 H to create recoil protons or by thermalizing and capturing these neutrons at thermal energies. Both approaches have drawbacks and are complex in deriving the energy spectrum through the unfolding process. The CLYC scintillator uses a distinct proton peak, whereby the position on the spectrum is proportional to the energy of the incident neutron. The response function of this detector has been simulated using Monte Carlo N-Particle eXtended code (MCNPX) for gamma-rays and neutrons of different energies. The obtained data has been discussed and analyzed.