Neutron Flux Levels Research Articles

Advancing Neutron Detection: Fabrication, Characterization, and Performance Evaluation of Self‐Powered PIN BGaN/GaN Superlattice‐Based Neutron Detectors

Solid state semiconductor based neutron detectors have the potential to be energy efficient and compact, making them suitable for applications where low power consumption and size constraints are important considerations. Herein, neutron detection devices based on PIN structures consisting of BGaN/GaN superlattice (SL) are demonstrated. These SL structures enable to incorporate significant boron (B) content and achieve good crystalline quality epilayers crucial for better neutron detection. Further, by leveraging the built‐in electric field generated by the PIN structure, these devices can be operated without any applied bias, simplifying overall operation and enabling a more compact size system for detection. Their performance is evaluated by measuring real‐time current response under neutron irradiation (IN) and without it (ID). The neutron induced current density (ΔJ = JN − JD) is determined, reaching an impressive value of 0.67 pA cm−2 (two times JD) under thermal neutron flux of 1.2 × 104 n cm−2 s−1 without biasing, demonstrating their self‐powered capability. They exhibit a linear response to varying thermal neutron flux levels. Additionally, the detectors successfully detect low thermal neutron fluxes down to 300 n cm−2 s−1, showcasing their potential for diverse applications, including in low neutron environments, screening nuclear warheads, and preventing illegal trafficking of radiological materials.

Radwaste analysis and package optimization with the F4E-radwaste tool and application to the ITER IVVS

Radwaste analysis and package optimization with the F4E-radwaste tool and application to the ITER IVVS

Effective Thermal Properties of a HTS Transition Edge Bolometer for High‐Flux Neutron Detection

AbstractMany applications require monitoring of increasingly high neutron flux levels. Pre‐neutron characterization is performed of a superconducting transition edge bolometric device, sensitive to neutrons by an enriched boron carbide (10B4C) layer. Heat from the absorption of neutrons is simulated using an attenuated laser, while the detector is cooled to the critical temperature, Tc. Frequency dependent (4–25 Hz) and constant input power measurements are performed on a 10B4C‐coated and a non‐coated pixel. Results are used to fit a two lumped element thermal model. The effective specific heat is highly dependent on the input pulse duration due to the interconnected stack structure. For modulated pulses, the active volume in the thermal circuit is the full depth of the substrate in the area under the pixel. For constant power input, a large portion of the substrate is heated and the interface conductivities to the cold finger are found to be important. The detectivity is D* = 3.5 × 108 , and is only slightly lower than similar detectors. Understanding the dynamics of the detector better and using neutron excitation, it is expected that optimizations can lead to an equally good detector for cold to thermal neutrons at flux levels >106 n cm−2s−1.

Neutronic simulation of medical radioisotope 99Mo and 177Lu production in IPR 14 MeV neutron generator facility

An accelerator based 14MeV neutron generator is commissioned at Institute for Plasma Research India. The generator is based on the linear accelerator concept where the deuterium ion beam impinged to the tritium target to produce neutrons. The generator is designed to produce 1×1012 neutrons per sec. The 14MeV neutron source facilities are an emerging tool for the lab scale experiments and research. In order to utilize the generator for the welfare of humanity, the assessment is made for the production of medical radioisotopes using the neutron facility. The usage of radioisotopes in the treatment and diagnosis of a disease is an important factor in the healthcare sector. A series of calculations are conducted to generate radioisotopes, especially 99Mo and 177Lu those are having huge applications in the medical and pharmaceutical industries. 99Mo can be also generated through neutron reactions 98Mo(n, g)99Mo and 100Mo(n, 2n)99Mo apart from fission reaction. The cross section of 98Mo(n, g)99Mo is high in the thermal energy range whereas 100Mo(n,2n)99Mo occurs at a high energy range. 177Lu can be produced using the reactions 176Lu (n, g)177Lu and 176Yb (n, g)177Yb. The cross section of both 177Lu production routes is higher at thermal energy range. The neutron flux level near the target is around 1010cm-2s-1. In order to enhance production capabilities, the neutron energy spectrum moderators are used to thermalize the neutrons. The materials used as a moderator are beryllium, HDPE, graphite, etc. Moderators enhance the capabilities of medical isotope production in neutron generators.

Research of the spatial effect of PWR fuel pin with disperse particles

• The spatial effect of the double-heterogeneous FCM fuel in PWR is concluded. • The spatial effect of PWR fuel pin embedded with disperse BPPs is studied. • Sensitivity analysis on the packing fraction and size of kernel are performed. • Relevant neutronic parameters are detailly discussed and synthetically analyzed. Fully ceramic microencapsulated (FCM) fuel is one of the most popular accident tolerant fuels (ATF) and it has a particular spatial structure, made of SiC matrix embedded with TRISO fuel particles. Fuel pin embedded with disperse burnable poison particles, has similar spatial structure. They are both double-heterogeneous and they have much unique neutronic characteristic. In this paper, efforts are made to analyze how the spatial structure of the disperse fuel particles and burnable poison particles affects their neutronic behaviors in detail, from pin-cell level, which is also referred as spatial effect. Many important neutronic parameters are discussed, including infinite multiplication factor (Kinf), four factors (fast fission factor ∊, resonance escape probability factor p, thermal utilization factor f and thermal reproduction factor η), neutron flux level, nuclides number density. They are compared with an equivalent homogeneous structure and the corresponding differences would reveal the unique spatial effect of this kind of double-heterogeneous structure. Besides, sub-regions analyses are also carried on. Finally, it is proved that the double-heterogeneous spatial structure would bring much influence on neutronic behaviors, and the spatial effect of disperse burnable poison particles would be even more serious.

Validation of fast neutron reactors fertile blanket depletion calculations through the analysis of the DOUBLON experiment in Phenix with TRIPOLI-4® and DARWIN3-SFR

• DARWIN3-SFR aims to be the new reference at CEA for deterministic fuel depletion calculations. • We analyze measurements from the DOUBLON experiment in the fertile blanket of the PHENIX reactor. • The DARWIN3-SFR results are also compared to its predecessor DARWIN-2 and the reference stochastic code TRIPOLI-4. • The calculations reproduce well the average neutron flux and the production of most nuclides in the fertile blanket. • Important spatial discrepancies are observed and the production of 241 Pu and 242 Pu is not well reproduced. A reliable assessment of the final inventory in the fertile blanket of a Fast Neutron Reactor is an important issue for fuel cycle physics, as it impacts safety, reprocessing and design studies. The performances of fertile blankets neutronics calculations is a long standing issue, as there as specificities in these regions that are quite challenging, especially for depletion codes. The new CEA fuel depletion calculation package is DARWIN3-SFR, which incorporates the deterministic neutronics code APOLLO3® and the depletion module MENDEL. This paper details the validation of DARWIN3-SFR for fertile blanket calculations through the re-interpretation of the DOUBLON pin-irradiation in the Phenix reactor. During this irradiation, nine pins are studied through isotopic ratios, which give us information about the depletion and, indirectly, the neutron flux calculations inside the blanket. This environment is especially challenging for neutronics codes, since there is a strong variation of the neutron energy and population over a short distance. Our recent analysis of TRAPU - which is a similar experiment, but in the core center - has proven DARWIN3-SFR to be reliable for the fuel depletion calculations of fissile subassemblies; nevertheless, its performances in the fertile blankets still require validation. We observe that, once the calculated neutron flux level has been adjusted to the experiment through the 148 Nd/ 238 U ratio, DARWIN3-SFR provides results similar to the reference stochastic code TRIPOLI-4®. However, both codes have difficulties to reproduce some of the measured isotopic ratios inside the fertile blanket. Whilst DARWIN3-SFR produces identical results for most of the isotopes analysed ( 234 U, 235 U, 236 U, 238 Pu, 239 Pu, 240 Pu), variations of the neutron spectrum lead to some disparities for the production of 241 Pu and 242 Pu. Indeed, the low-energy flux estimation is higher with TRIPOLI-4® than with DARWIN3-SFR in the energy range where the 240 Pu has a high capture cross section, which increases the calculated production of 241 Pu and 242 Pu. DARWIN3-SFR and its predecessor DARWIN-2 are efficient at calculating the average neutron flux level over the entire fertile blanket. However, the results of the two codes show a strong pin-to-pin dispersion, resulting in a different shape of the neutron flux inside the blanket. With DARWIN3-SFR, the estimated neutron spectrum is softer than in DARWIN-2, which impacts the 238 U capture and fission reaction rates and hence the production of plutonium.

Design and neutronic analysis of the intermediate heat exchanger of a fast-spectrum molten salt reactor

Various research groups and private interprises are pursuing the design of a Molten Salt Reactor (MSR) as one of the Generation-IV concepts. In the current work a fast neutron MSR using chloride fuel is analyzed, specially analyzing the power production and neutron flux level in the Intermediate Heat Exchanger (IHX). The neutronic analysis in this work is based on a chloride-fuel MSR with 600 MW thermal power. The core power density was set to 100 MW m −3 with a core H/D [[EQUATION]] 1.0 amd four Intermediate Heat Exchanger (IHX). This leads to a power of 150 MW per IHX; this power is also comparable to the IHX proposed in the SAMOFAR framework. In this work, a preliminary design of a 150 MW helical-coil IHX for a chloride-fueled MSR is prepared and the fission rate, capture rate, and inelastic scatter rate are evaluated.

Reactor Pulse Operation for Nuclear Instrumentation Detector Testing – Preparation of a Dedicated Experimental Campaign at the JSI TRIGA Reactor

The availability of neutron fields with a high neutron flux, suitable for irradiation testing of nuclear instrumentation detectors relevant for applications in nuclear facilities such as material testing reactors (MTRs), nuclear power reactors and future fusion reactors is becoming increasingly limited. Over the last several years there has been increased interest in the experimental capabilities of the 250 kW Jožef Stefan Institute (JSI) TRIGA research reactor for such applications, however, the maximal achievable neutron flux in steady-state operation mode falls short of MTR-relevant conditions. The JSI TRIGA reactor can also operate in pulse mode, with a maximal achievable peak power of approximately 1 GW, for a duration of a few ms. A collaboration project between the JSI and the French Atomic and Alternative Energy Commission (CEA) was initiated to investigate absolute neutron flux measurements at very high neutron flux levels in reactor pulse operation. Such measurements will be made possible by special CEA-developed miniature fission chambers and modern data acquisition systems, supported by the JSI TRIGA instrumentation and activation dosimetry. Additionally, measurements of the intensity of Cherenkov light are proposed and being investigated as an alternative experimental method. This paper presents the preparatory activities for an exhaustive experimental campaign, which were carried out in 2019-2020, consisting of test measurements with not fully appropriate fission chambers, activation dosimetry and silicon photomultipliers (SiPMs) The presented results provide useful and promising experimental indications relevant for the design of the experimental campaign.

Development of a Wideband Current Amplifier dedicated to Fission Chamber Measurement

Fission chambers are widely used in nuclear reactors, either occasionally to probe the neutron flux maps in nuclear power plants or as part of the nuclear instrumentation used in research reactors to characterize neutron flux levels in irradiation locations. Wide dynamic range acquisition systems are very limited in the nuclear instrumentation market. Thus, since 2012, CEA is developing the MONACO system, which allows measurement in pulse, fluctuation, and current mode running simultaneously on the same channel. The simultaneous use of three measurement modes places high requirements on the amplification stage. Indeed, the current mode exploits a bandwidth between 0 and a few kHz, while the fluctuation mode is based on the frequency range of 100 kHz 1MHz. Finally, for the pulse mode, it is necessary to identify current pulses of a few uA in amplitude and with a duration of about ten nanoseconds, which implies a bandwidth that extends up to 10 or 20 MHz. In the frame of the industrialization of the MONACO system, a new preamplifier was designed to meet those requirements. With a first prototype, LDCI and I-TECH performed very promising bench tests in October 2020. Tests with synthetic signals corresponding to pulse and fluctuation mode showed very good performance of the electronics. The positive and negative current pulses are correctly amplified and the measurement noise was estimated at 0.1 µApp thanks to proper electronic shielding. The DC measurement is also satisfactory in the tested range with an accuracy of less than nA and a bandwidth of 1kHz. Preamplifier qualifications are planned within realistic experiments using a fission chamber placed in a neutron field.

Light Yield Response of Neutron Scintillation Screens to Sudden Flux Changes.

We performed a study of the initial and long term light yield of different scintillation screen mixtures for neutron imaging during constant neutron irradiation. We evaluated the light yield during different neutron flux levels as well as at different temperatures. As high frame rate imaging is a topic of interest in the neutron imaging community, the decay characteristics of scintillation screens are of interest as well. Hence, we also present and discuss the decay behavior of the different scintillation screen mixtures on a time scale of seconds. We have found that the decay time of ZnS:Cu/6LiF excited with a high neutron flux is potentially much longer than typically stated. While most of the tested scintillation screens do not provide a significant improvement over currently used scintillation screen materials, Zn(Cd)S:Ag/6LiF seems to be a good candidate for high frame rate imaging due to its high light yield, long-term stability as well as fast decay compared to the other evaluated scintillation screens.

Characterization of the neutron spectra in three irradiation channels of the JSI TRIGA reactor using the GRUPINT spectrum adjustment code

Reference neutron fields in irradiation facilities are crucial for experimental research work in nuclear science and technology in general, and at the fundamental level for the measurement of neutron cross sections and validation of evaluated cross-section data. Monte Carlo particle transport techniques in conjunction with state-of-the-art nuclear data libraries are becoming the reference in the broad context of nuclear reactor analysis, and are extensively used for detailed computational characterization of irradiation facilities, including the calculation of neutron spectra. However, adjustment of the neutron spectra based on experimental data is still required. In this work we present the characterization of the neutron spectra in the three most important irradiation channels of the JSI TRIGA reactor with different spectral characteristics, as an important step in developing the experimental capabilities of the reactor for the purpose of nuclear data validation through activation dosimetry measurements. In the first part of this work we performed detailed Monte Carlo calculations of the neutron spectra, comparing the results obtained using different nuclear data libraries. In the main part of this work we employed the JSI developed GRUPINT spectrum adjustment code for the characterization of the neutron spectra, on the basis of the calculated neutron spectra and measured reaction rate ratios. A comprehensive overview of the methodology employed in the code and its abilities is given. We observe that a positive compensation of up to 10 % in the thermal part of the spectrum is needed, while no compensation is needed in the epithermal and fast parts. Absolute neutron flux levels were computed from the adjusted neutron spectra and absolute reaction rate measurements, and compared to the results of a previous analysis using only the Monte Carlo calculations. The work presented in this paper represents the basis for future experimental and analytical work, in particular extending the scope of the measured capture and threshold dosimetry nuclear reactions, to support a thorough characterization of the neutron spectra in the irradiation channels of the reactor. This effort will enable the JSI TRIGA reactor to support a broad scope of experiments for fundamental research as well as development or testing activities at the highest level.

Comparison Between Silicon Carbide and Diamond for Thermal Neutron Detection at Room Temperature

Neutron radiation detector for nuclear reactor applications plays an important role in getting information about the actual neutron flux. Such a detector must be able to operate at high neutron flux levels (>109 cm−2 s−1) and discriminate the neutron and gamma responses in the nuclear reactor’s mixed neutron–gamma environment. Silicon carbide and diamond are the most attractive semiconductor materials for neutron detection, thanks to their outstanding properties, such as high displacement threshold energy and wide bandgap energy, which allow them to operate in high radiation levels and high temperature. The aim of this article is to compare the ability to detect thermal neutrons of these two semiconductors at the same irradiation conditions. For this purpose, the neutron irradiation tests of detectors were implemented at MINERVE research reactor at CEA Cadarache. The 4H-silicon carbide (SiC) p+n diode has demonstrated better neutron–gamma discrimination at 0-V bias voltage than at −200 V which is explained by its increased sensitivity to gamma photons at −200 V caused by a wider charge collection region than at 0 V. Therefore, it is preferable to use the 4H-SiC p+n diode without an external electric field for applications in the mixed neutron–gamma environment such as nuclear reactor environment. The results show that the single-crystal chemical vapor-deposited (sCVD) diamond-based detector has better neutron to gamma discrimination, thanks to the use of 6Li as a neutron converter instead of 10B. However, the study of the radiation stability of detectors showed that the sCVD diamond-based detector suffers from the “polarization effect” when it operates at a high neutron flux (~109 cm $^{-2}\,\,\cdot \,\,\text{s}^{-1}$ ).

Measurement of the Gas Velocity in a Water-Air Mixture in CROCUS Using Neutron Noise Techniques

The possibility of measuring the gas-phase velocity in a two-phase mixture through the use of neutron noise techniques is demonstrated in the zero-power reactor CROCUS of the Ecole Polytechnique Federale de Lausanne. It is the first step toward the experimental validation of an existing theoretical model whose objective is the reconstruction of the void profile in a channel. The use of zero-power research reactors is advantageous due to their clean environment in terms of signal fluctuations. To this end, a channel was installed in the reflector of CROCUS. A two-component mixture is generated inside the channel through the injection of compressed air. The signal fluctuations of neutron detectors located at various axial locations next to the channel are processed to determine the transit time of the gas phase between detectors. Four methods are presented based on the detector signal time series either in the time domain (time correlations between signals) or in the frequency domain (phase of the cross-power spectral density. All four methods returned consistent transit times and similar experimental uncertainty. The largest possible gas injection rates as well as the highest possible neutron flux level improve the visibility of the traveling perturbation and reduce the experimental uncertainty on the transit time for a given acquisition time.

The study for activation evaluation on activated structures in nuclear power plant with 40 years operation history

Abstract Reactor pressure vessel internal, reactor pressure vessel and bio-shield are one of the activated structures in nuclear power plant. Especially, reactor pressure vessel internal which is irradiated by neutrons for several decade of years to the level of 1E + 11 ∼ 1E + 13 neutrons/cm2.sec is highly activated one and specific activity for some nuclide is to the level of ∼1E + 09 Bq/g. In case of reactor pressure vessel, neutron flux level is 1E + 09 ∼ 1E + 11 neutrons/cm2.sec and specific activity is to the level of ∼1E + 06 Bq/g, so, it could be classified as low activated region. Bio-shield can be divided into two parts – very low activated and clearance regions, and neutron flux and specific activity in very low activated region are ∼1E + 08 neutrons/cm2.sec and ∼1E + 01 Bq/g. Separation between two regions in bio-shield is determined by the factors of impurities’ kind and quantity included in it. Major activated nuclides in those structures are Ni-63, Fe-55, Co-60, Ni-59, Mn-54, H-3, C-14, etc. through neutron capture reaction – (n, γ), (n, p), (n, 2n), and (n, α). Activation evaluation for structures in nuclear power plant is performed using MCNP and ORIGEN codes.

Reactor physics evaluation of the TRIGA LEU fuel in the 20 MW NIST research reactor

Abstract This paper performs a neutronics evaluation of the General Atomics UZrHx low enriched uranium fuel – TRIGA fuel – in the National Bureau of Standards Reactor (NBSR) at the National Institute of Standards and Technology (NIST). The objective of this study is to examine the accountability and sustainability of the TRIGA fuel on neutronics aspects when applying it to the NBSR conversion. A feasibility scoping study was previously undertaken with considerations on various fuel dimensions, fuel rod layout configurations, and structure material selections, identifying the best option of deploying the TRIGA fuel to NBSR. Continuing with these efforts, an equilibrium NBSR core using the identified fuel was generated, and a well-round physics assessment was carried out by examining key neutronics performance characteristics of the core. All calculations were completed with MCNP-6, a 3-D Monte Carlo neutron transport code. The same fuel management scheme and fuel cycle length as the existing NBSR was adopted in the equilibrium core generation adopts to retain performance consistencies. The effectiveness of the fuel was examined at four representative burnup states of the fuel cycle. Neutronics performances of the equilibrium core was characterized by the fast and thermal neutron flux level as well as power distribution in the core. Reactor safety related parameters such as kinetics parameters and power peaking factors were also evaluated in the study. All results were compared against the current NBSR fueled with HEU for justifications. The findings in this research prove the viability of the TRIGA fuel for the NBSR conversion, and provide supporting data for future investigations on this subject.

The investigation and calculation of the transmutation paths for the production of 252cf in fast reactors

Abstract The generation and transmutation process of the transplutoniums in the conversion chains for 252cf production is calculated and analyzed. The production and conversion efficiency of the chains with different target nuclides and neutron spectrums including fast, thermal and moderated neutron spectrums is computed. The conversion paths for 252cf under different neutron spectrums and neutron flux levels are analyzed. The multi-group depletion program STEP1.0 based on linear chain method is used to calculate the nuclide densities. The model of the reactor is simulated by the transport-burnup coupling system based on the STEP1.0 and MCMG-II for the detailed description of the target designs in the fast reactor. The differences of the conversion efficiency caused by the neutron spectrums can be qualified more accurately with the multi-group cross section calculations. The higher conversion efficiency than the typical thermal neutron spectrum can be obtained by some moderated spectrum optimized for absorption reaction in fast reactor (CEFR China Experimental Fast Reactor) through the irradiation of some transplutoniums (242Pu, 241Am, 244Cm). The neutron flux of large scale demonstration fast reactor or prototype fast breeder reactor is usually higher than the experimental fast reactor (CEFR) or reactor device. China is also developing the demonstration fast reactor (CFR China Fast Reactor) which will have a maximum neutron flux higher than 5.0E15 n/(cm2·s). Therefore, the 252cf production will be more practical and efficient with CFR.

Simulation Study on the Wide-Range Measuring Algorithm in ITER Neutron Flux Monitor

The neutron flux monitor (NFM) is one of the most important diagnostic systems for ITER. Wide-range measuring algorithm (WRMA) is the core algorithm in the NFM system, which deals with the key task of neutron flux measurement. In this paper, the principle and implementation of WRMA, including counting and Campbelling algorithms, are introduced in detail, with error sources of the two algorithms analyzed. In order to study the performance of WRMA, we established a simulation system for neutron signal processing using MATLAB. According to the principle of neutron pulse distribution, the digital waveforms at different neutron flux levels were simulated as inputs to the WRMA module. The variation of measuring error was studied by comparing the counting and Campbelling results with actual input counting rate. In addition, the effects of different neutron pulse widths on the results of the algorithm were simulated. A preliminary experiment at HL-2A was carried out to validate the algorithm.

In-Pile Qualification of a Fast-Neutron-Detection-System

In order to ensure the quality and the relevance of irradiation programs in the future Jules Horowitz Reactor (JHR), the French Alternative Energies and Atomic Energy Commission (CEA) has significantly increased its research and development effort in the field of in-pile instrumentation during the last decade. Major progress has been achieved in the capability to perform accurate in-pile measurements using reliable and updated techniques. A significant part of this effort has been conducted in the framework of the Joint Instrumentation Laboratory between the CEA and the Belgian Nuclear Research Centre (SCK $\cdot$ CEN). In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile. The fast-neutron-detection-system (FNDS) has been designed to monitor accurately high-energy neutron flux (E > 1 MeV) in typical material testing reactor (MTR) conditions, where neutron flux levels can be as high as $10^{15}\,\,\text {N}\cdot \text {cm}^{-2}\cdot \text {s}^{-1}$ and are generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few $10^{15} \gamma \cdot \text {cm}^{-2}\cdot \text {s}^{-1}$ , which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors allowing the simultaneous detection of both thermal- and fast-neutron flux. Thermal neutrons can be measured using a self-powered neutron detector or a 235U miniature fission chamber, while fast neutron detection requires a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long-term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK $\cdot$ CEN in Belgium. Two FNDS prototypes were operated in-pile during nearly 1000 h. These tests exhibited the consistency of the measurement of thermal to fast neutron flux ratio with Monte Carlo neutron and photon transport calculations, as well as the right compensation of fissile material depletion. This paper describes a second test recently completed at ISIS reactor operated by CEA in France. For this irradiation, FNDS signal was compared to reference thermal- and fast-neutron flux measurements using activation dosimeters analyzed under COFRAC Quality Certification. During this latter test, FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%. FNDS is now operational and is assumed to be the first and unique acquisition system able to provide an online measurement of the fast neutron flux in MTR conditions. This system will be used to perform spectral neutron characterization of JHR channels, but it may also be implemented in future irradiation experiments, for a better and real-time evaluation of the fast neutron flux received by material and fuel samples.

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

In-Pile Qualification of the Fast-Neutron-Detection-System

In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile by the French Alternative Energies and Atomic Energy Commission (CEA) in cooperation with the Belgian Nuclear Research Centre (SCK•ECEN). The Fast-Neutron-Detection-System (FNDS) has been designed to monitor accurately high-energy neutrons flux (E > 1 MeV) in typical Material Testing Reactor conditions, where overall neutron flux level can be as high as 1015 n.cm-2.s-1 and is generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few 1015 γ.cm-2.s-1, which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors, including a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK•CEN in Belgium. Then a second test was recently completed at ISIS reactor operated by CEA in France. FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%.