Fast Neutron Fluence Research Articles

Determination of the elemental composition and changes in thermal and electrical conductivity of “Hanford” and low-ash medium-grained graphite’s grade graphites depending on the fast neutrons fluence

Studying the properties of graphite recovered from operating nuclear reactors is vital for predicting the properties and integrity of graphite as part of assessing the continued operation and life extension of nuclear reactors. The purpose of the study is to determine the electrical conductivity and thermal conductivity of low-ash medium-grained graphite’s grade graphite in the thermal column masonry of the VVR-SM research reactor in the measurement temperature range corresponding to the conditions of normal operation of the reactor to determine the service life. In this work, the change in thermal conductivity and electrical conductivity of GMZ graphite was studied, as well as for comparison of Hanford grade graphite depending on the fluence of fast neutrons and the measurement temperature. The dependence of electrical conductivity and thermal conductivity on dose and temperature has been established. It has been shown that the greater the neutron fluence, the more both the thermal conductivity and electrical conductivity of the material decreases. The service life of the thermal column has been determined.

Dopant, coating, and grating effects in silica optical fibers under extreme neutron irradiation

Radiation effects on fiber Bragg gratings (FBGs) have been studied but data gaps related to the effects of displacement damage resulting from high fast neutron fluence remain. In this work, Type-I and Type-II FBGs inscribed in optical fibers with various core dopants (Ge and F) and fiber coatings (acrylate, polyimide) were monitored in situ during 75 days of neutron irradiation to a peak fast (>0.1 MeV) neutron fluence of 3×1021 nfast/cm2. The reflected intensity of the Type-I FBGs inscribed in a Ge-doped core fiber decreased by >30 dB within 7 h of irradiation (1019 nfast/cm2), whereas Type-II FBGs inscribed in a pure silica core fiber eventually approached >40 dB attenuation after accumulating a fast neutron fluence on the order of 1020 nfast/cm2. Type-II FBGs inscribed in F-doped core fiber improved stability: the attenuation approached an equilibrium value in the range of 10 to 20 dB.

Optical spectra in pure-silica-core single mode optical fibers after high-fluence reactor irradiation

The paper describes the study of radiation-induced attenuation (RIA) spectra in reactor-irradiated single mode at λ=1.55 μm optical fibers (OFs) with differente high-temperature coatings (polyimyde, aluminum and copper). The OFs were previously irradiated up to fast neutron fluence 1.8·1020 n/cm2 and gamma dose 2.32 GGy. After three years of storage at room temperature (RT), the RIA spectra were examined in the range of 650–1700 nm. All RIA mechanisms acting in this spectral range were identified: a tail of non-bridging oxygen hole center, a long-wavelength RIA (LWRIA) band with a maximum at λ > 1600 nm, unstructured “grey” loss due to coating microbendings or damage, and an absorption band peaking at 930 nm (E=1.33 eV). It was found out that the LWRIA can be significantly annealed at RT in polyimide-coated OFs. Grey loss were found to be more stable than those causing LWRIA.The LWRIA arising in reactor- and gamma-irradiated OFs have close thermal stability at RT and exhibit a similar shapes, which close to the absorption band of self-trapped holes (STH). This suggests a STH-like nature of the LWRIA.

Materials Science Analysis of RCCA Structural Materials and Calculations for Lifetime Extension

The paper considers the possibility of lifetime extending of the rod control cluster assemblies in operation by determining their residual lifetime using an approach that takes into account the actual position of the rod control cluster assembly in the core. Excessive conservatism was introduced by assuming that the immersion of the 10th working group in the core during all fuel campaigns is 70 %, while the actual position is 85-90 %. Taking into account the actual position of the rod control cluster assemblies indicates much less irradiation of the lower, most loaded part of control rod cladding and much less burnup of the 10B compared to 70 % immersion of the rod control cluster assemblies, since they are in the area of lower neutron flux density.
 The results of a comprehensive analysis of available studies and test results to determine the phenomena that can potentially approach the achievement of the boundary state of the structural components or limit the lifetime of control rods and rod control cluster assemblies in general are presented. Such phenomena include general corrosion of structural materials (42CrNiMo, Cr18Ni10Ti) (decrease in wall thickness, increased transfer of radioactive corrosion products the coolant); radiation resistance (degradation of mechanical characteristics) of control rod cladding materials (42CrNiMo) and other structural components of the rod control cluster assemblies (42CrNiMo, CrKh18Ni10Ti); radiation resistance of neutron-absorbing materials (B4C, Dy2O3∙TiO2); compatibility of absorbing materials with the control rod cladding.
 Based on the results of post-irradiation examinations of control rods tested in research reactors as dummy rods or rod control cluster assembly rods operated in the VVER-1000 core, the following boundary values were established that can be used as criteria for maintaining the integrity of the control rod cladding: maximum fast neutron fluence in the lower, most loaded part of the control rod cladding (in the weld area of the lower end plug and the cladding) should not exceed 34×1021 n/cm2; maximum burnup of the 10B isotope should not exceed 45 %. If these two criteria are met, the possibility of other potentially damaging phenomena is too low.
 According to the calculations that use the schedule of changes in the position of the rod control cluster assemblies of the control group and the power of one of the power units during one fuel cycle, one can preliminary conclude that the proposed approach will extend the lifetime of almost all rod control cluster assemblies from 25500 to 38000 hours in the automatic control group and from 75600 to 113500 hours in the shutdown group. The residual lifetime should be assessed for a particular power unit and each rod control cluster assembly separately, since each rod control cluster assembly has its own operating history.

Prediction of radiation embrittlement of VVER-1000 RPV welded seam material at lifetime extension up to 60 years and more

The correctness of the model assumption about additivity of the contributions of hardening (formation of dislocation loops and radiation-induced precipitates) and non-hardening (formation and accumulation of grain boundary segregations) mechanisms to the final radiation embrittlement of VVER-1000 RPV weld materials has been verified. The paper presents the results of the experiment to obtain predicted values of the critical embrittlement temperature of weld metal with nickel content of 1.59–1.88% using accelerated irradiation in the IR-8 research reactor from states characterized by different starting level of grain boundary segregation. The procedure of accounting for the flux effect for the values of critical embrittlement temperature obtained using accelerated irradiation has been proposed. The conservativity of the weld metal radiation embrittlement dependence has been verified based on the modal assumption of additivity of the contributions of the strengthening and non-strengthening mechanisms up to the values of fast neutron fluence ~75–1022 neutrons/m2.

Estimation of the Fast-Neutron Fluence Rate With a 4H-SiC-Based Detector in a 14.1-MeV Mono-Energetic Field

Estimation of the Fast-Neutron Fluence Rate With a 4H-SiC-Based Detector in a 14.1-MeV Mono-Energetic Field

Characterization of mineral insulated cables at the WWR-K reactor: First results

At the WWR-K reactor, studies were started on the radiation resistance of signal cables with two types of mineral insulation (MgO and Al2O3). It is planned to accumulate fluence of fast neutrons ∼ 1020 cm−2 in cables. The irradiation temperature will be 500 ± 50 °C. The in-situ study of the degradation of the electrical properties of the insulation of signal cables will be carried out. This paper presents a capsule design, the results of complex calculations for the development of the capsule design, the expected neutron fluences, the dpa in steel, the technique for in-reactor measurement of electrical characteristics, and a work plan for the further studies. It is shown that the developed capsule forms the necessary conditions for irradiation. In particular, the temperature control range is 400–600 °C. The cable irradiation time until the target neutron fluence is reached will be about 100 effective days.

The fast neutron fluence and the activation monitor activity calculations using fine multigroup and continuous nuclear data

The fast neutron fluence and the activation monitor activities are routinely calculated with TORT deterministic code and BUGLE-B7 nuclear data library with 47 broad energy groups. The objective of the paper is to analyse options to improve reactor dosimetry transport calculations. There are two paths to improve reactor dosimetry calculations. Increasing geometry, angular and energy mesh size is applicable for TORT code while using newer nuclear data libraries is relevant for both deterministic and Monte Carlo codes. Two new calculation options (improved TORT and Monte Carlo MCNP6) were compared with the standard TORT calculation for VVER-440 Dukovany Unit 3 Cycle 31. The fast neutron fluence with 0.5 MeV threshold as well as activity of Fe. Ni, Ti, Cu. Mn and Nb monitors were evaluated. Standard TORT calculations were improved from S16P3 to S30P3 with three times finer axial mesh size. 120° core symmetry r-ϑ mesh size with 0.5° step and fine multigroup libraries VUAMIN-B7 with 199 neutron energy groups and ENDF/B-VH.1 with 200 neutron energy groups. Both ENDFB and IRDFF activation cross sections were used. The drawback of expanded mesh size is raised calculation runtime since TORT deterministic code is not parallelized and one calculation can require multiple weeks of CPU time. An alternative option of using MCNP6 Monte Carlo code with continuous ENDF/B-VH.1 nuclear data with detailed 3-D geometry and pin-wise effective neutron source prepared by MOBY-DICK diffusion code reactor analysis wras explored. It was found that using finer mesh size affects reactor dosimetry’ tallies less than the choice of nuclear data library. BUGLE-B7 and VTTAMIN-B7 produce results typically within 1% difference. ENDF/B-VH.1 calculations with 200 neutron energy’ groups with TORT code are even in better agreement with MCNP6 calculations with continuous nuclear data libraries. The largest differences of around 2% were observed between VTTAMIN-B7 library based onENDFB-VH.O nuclear data and ENDF/B-VH.1 library. Nuclear data library’ has larger impact on the results with up to 7 % difference between all 0.5 MeV fast neutron fluence calculations. The largest intact of nuclear data was observed for Mn(n.2n) monitor.

Calculation and analysis of fast neutron fluence of pressure vessel in AP1000 reactor

The structural integrity of the reactor pressure vessel plays an important role in the safe operation of AP1000 nuclear power plant. Radiation damage of fast neutrons is the main factor leading to pressure vessel material deterioration. Therefore, it is necessary to accurately calculate the fast neutron fluence in the pressure vessel. Discrete ordmates method radiation shielding analysis code cosSN is used to calculate fast neutron fluence hi the pressure vessel and the calculation result is compare to measurement of actual AP1000 power plant. Bondarenko method is adopted to calculate multi-group cross section hi resonance energy region, and Bell-Hansen-Sandmeier(BHS) transport correct method is used to deal with anisotropic scattering effect. To improve computational efficiency, cosSN adopts the diffuse comprehensive acceleration and the multi-grid technique to accelerate source iteration, and adopts parallel computing to solve the storage limitations in large-scale calculations. For fast neutron fluence calculation, the calculation uncertainty conies from input parameters and approximation of theoretical model. The results hidicate that the calculated specific activity is conform to the actual measured value of the power plant. The calculation uncertainty is ± 13.28%. which meets the requirements of China domestic energy industry standards NB/T 20576-2019 and NRC RG 1.190.

A quantified approach to assessing the effectiveness of nuclear reactor pressure vessel neutron fluence monitoring programs

This paper presents a quantitative approach to characterize the effectiveness of neutron fluence monitoring programs using adjoint neutron transport calculations. The results of adjoint calculations are useful for this purpose because the adjoint solution is the expected contribution from one region of phase space to a response of interest, i.e., the importance of that region of phase space to the response. Thus, by identifying a response of interest (for this work, the fast (E > 1.0 MeV) neutron fluence rate at the peak fluence rate location in the reactor pressure vessel (RPV)), one can identify the fuel regions most important to that response. Once identified, the ability of various surveillance programs to adequately monitor the most-important fuel region(s) can be assessed. The approach described is applied to a three-loop Westinghouse pressurized water reactor to compare the effectiveness of surveillance-capsule neutron dosimetry versus ex-vessel neutron dosimetry to monitor the fuel assemblies that contribute to the peak fast (E > 1.0 MeV) neutron fluence rate in the RPV.

On the effect of the core boundary power distribution on the radiation situation in reactor components

The assessment of the neutron flux distribution in nuclear power plant components, like reactor internals and the pressure vessel, is one of the most important parts of their residual lifetime evaluation process because the irradiation damage of these components is strongly dependent on it. The fast neutron fluences in power reactors are generally determined using calculations and verified by measurements. Discrepancies between them sometimes occur, which can be caused by be the inaccuracies in the power distribution used in neutron transport calculations. This paper quantifies the effect of uncertainty in power density on the neutron fluences behind the reactor pressure vessel (RPV). An increase in power density was experimentally simulated by implementing a higher uranium enrichment of certain selected pins. The experiment was also simulated using the MCNP code with the ENDF/B-VII.1 library. Both the experimental and calculational data show a significant local increase in neutron flux. Even behind the RPV, the increase is as high as 25%, while the local power increase in the perturbed pins is about 70–80%, which is approximately equal to the 2 σ power density uncertainties. A good agreement between measurement and calculation was found.

Transmutation detectors for fast neutron fluence measurements

Transmutation detectors (TMDs) are based on a similar principle as neutron activation detectors (NADs). In both cases, the neutron fluence is estimated from the number of reactions in the detectors. The main differences between these two detectors are the type of product created by the nuclear reaction and the evaluation of the irradiated detectors. In TMD, the transmuted nuclide is stable, and the number of reactions is gained by analytical methods suitable for measuring low concentrations. In NAD, the nuclear reaction leads to the production of radionuclides and the number of reactions is evaluated from the activity measured by radiometric instruments. While TMDs for thermal neutrons can be easily prepared and evaluated due to their high thermal neutron cross sections for isotopes such as 10B, 115In or 197Au, the requirements for fast neutron TMDs are higher. Cross sections for nuclear reactions suitable for the TMD method with fast neutrons are in the range of 1 to 100 mb for the fission spectrum. This leads to concentrations from 0.01 ppm to 1 ppm for fast neutron fluence of 1E19 cm-2. These low concentrations imply requirements for detector material purity and sensitivity of concentration evaluation equipment. Materials for fast neutron TMDs were proposed, where the relevant parameters, such as energy threshold and minimal detectable fluence are known and convenient. An experimental comparison of some TMD and NAD responses was made in the LVR-15 research reactor using secondary ion mass spectrometry to evaluate the concentration of TMDs.

Westinghouse fluence methodology uncertainty estimates for reactor vessel internals fast neutron exposure evaluation in subsequent license renewal period

The United States Nuclear Regulatory Commission (USNRC) has expressed concerns about the calculated fast neutron fluence exposure uncertainties for the reactor vessel internals (RVI) of light water reactors for extended operation [1, 2]. The current guidance for calculating neutron fluence, Regulatory Guide (RG) 1.190 [3], was developed with a focus on the reactor pressure vessel (RPV) region surrounding the active core, which may not be adequate for the RVI above and below the active core during the license renewal periods, especially the subsequent license renewal (SLR) period. An updated version of ANSI/ANS 19-10 [4] may fill the regulatory guidance gap on neutron fluence determination methodology for RVI but still faces the challenges of lacking RVI fluence measurement benchmark data and analytical uncertainty analysis, which are two cornerstones for qualifying a neutron fluence determination methodology. The Pressurized Water Reactor Owners Group (PWROG) has responded to this regulatory concern and have performed analytical uncertainty analysis and measurement benchmark work to address these two technical challenges for the nuclear power industry.

Ex-vessel neutron dosimetry results in the extended beltline region

Embrittlement of the pressure vessel has been a long-standing concern for pressurized water reactors. Historically, the beltline region of the pressure vessel, or the region of the pressure vessel located opposite the active fuel, has been the primary focus of embrittlement concerns. Embrittlement evaluations typically include a calculation of the neutron flux incident on the pressure vessel, which is in part validated with measurements. In the context of life extension, the beltline region, which is technically defined as the portion of the pressure vessel experiencing fast neutron fluence (E > 1.0 MeV) equal to or greater than 1017 n/cm2, is growing axially. Present fluence calculations for many plants indicate that the beltline region at the end of the 60 years of operation may extend to the bottom of the pressure vessel nozzles. Calculations in the extended beltline are difficult to validate as there are few dosimetry measurements available. At one Westinghouse 4-loop plant, ex-vessel neutron dosimetry sensor sets were irradiated in the extended beltline region. The measurements in the extended beltline region compare well with the discrete ordinates (RAPTOR-M3G) transport calculations and indicate the net uncertainty may be on the order of 30% in the extended beltline region.

Improved determination of fast neutron fluence onto the WWER pressure vessel metal surveillance specimens

The special methodology for the determination of the neutron fluences onto the surveillance specimens of the WWER-1000 reactor pressure vessel metal has developed by INR NASU specialists and is successfully applied. The methodology is based on the Monte-Carlo code that is used for neutron transport calculations to the locations of the surveillance specimens. The methodology improvement is described. The fundamentals of the technique of the calculation-experimental determination of the fast neutron fluences onto surveillance specimens and their errors are presented. It is shown a dosimetry experiment in a particular reactor is needed to obtain experimental data about irradiation conditions in the locations of the surveillance specimens.

Electrical conductivity of GMZ grade graphite as a function of fast neutron fluence

Electrical conductivity of GMZ grade graphite as a function of fast neutron fluence

Perovskite-based detector for reactor dosimetry monitoring

Hybrid halide perovskites have demonstrated significant efficiency in detecting a broad spectrum of high-energy radiation, including X-rays, gamma rays (γ-rays), and neutrons. Given the common occurrence of mixed radiation fields, we investigated the performance of a perovskite-based detector in a neutron-gamma mixed field. A large methylammonium lead tribromide (MAPbBr3) single crystal (SC) was synthesized via the oriented crystal-crystal intergrowth method. This SC was used to fabricate a gamma detector with carbon electrodes, which was tested in the CROCUS zero-power reactor cavity. The detector’s photocurrent response exhibited a strong correlation with known gamma dose rates, as measured by an ambient Berthold LB 112 gamma probe, facilitating the accurate conversion of photocurrent to dose rate. Notably, the device did not exhibit degradation under neutron radiation exposure. To further assess the impact of neutrons, X-ray diffraction and electron paramagnetic resonance analyses were performed on small MAPbBr3 SCs grown by inverse temperature crystallization. These SCs were irradiated within the CROCUS reactor core and by a Pu-Be neutron source at liquid nitrogen temperature. Our findings indicate that the perovskite material can withstand the nominal in-core operation conditions of the CROCUS reactor. Additionally, it endures irradiation at liquid nitrogen temperature, corresponding to a fast neutron fluence of approximately 1010 cm-2 and a gamma radiation dose of about 50 Gy, confirming only the temporary creation of defects. No signs of long-term deterioration were observed, suggesting a potential self-healing mechanism. This resilience positions perovskite SCs as viable candidates for in-core radiation detection, supporting the further development of miniaturized MAPbBr3 SC devices for such applications.

Development of HELIOS/BIPR/PARCS/MCNP6 Computation Route for WWER RPV Neutron Fluence Analysis and Validation Against Ex-Vessel Detector Measurement Data

A computational route was developed for precise calculation of fast neutron fluence on a WWER-type reactor pressure vessel (RPV). The method is based on the transfer of neutronics data from HELIOS-2 lattice calculations and nodal diffusion neutronics data (power, density, and temperature) from BIPR7.1 and PARCS 3.36/PATHS core calculations into a three-dimensional (pinwise axially distributed) fixed neutron source for modeling of transport of fast neutrons from the reactor core to the outer surface of the RPV using MCNP6.2. Validation of the proposed computational method was carried out based on comparative analysis of MCNP6.2-predicted and neutron dosimetry–measured reaction rates [54Fe(n,p)54Мn, 93Nb(n,nʹ)93mNb, and 58Ni(n,р)58Со] on the outer surface of the Armenian Nuclear Power Plant (ANPP) Unit 2 RPV. Validation revealed that the MCNP6.2-predicted fast neutron fluence results are very sensitive to the ENDF-B neutron data. Particularly, MCNP6.2 with ENDF/B-VIII.0 significantly underpredicts (20% to 30%) fast neutron fluence while using ENDF/B-VII.1 data overpredicts it. Adding revised beta-released evaluations of 54Fe, 56Fe, 57Fe, and 16O from the International Nuclear Data Evaluation Network (INDEN) to ENDF/B-VIII.0 allows one to obtain reasonable agreement with measurement results for all types of measured reaction rates.

Investigations of microstructure and tensile properties of neutron irradiated 6061 aluminum alloys

Effects of neutron irradiation on the microstructure and mechanical properties of CMRR (China Mianyang Research Reactor) structural material pure aluminum and 6061 alloy were studied by scanning/transmission electron microscopy (S/TEM) and tensile tests. The fast neutron (E > 0.1 MeV) fluences of 1.09–2.89 × 1022 n•cm−2 produced damage doses of 1.6–4.2 displacements per atom (dpa). Voids and faulted 1/3<111> dislocation loops were observed in both pure aluminum and 6061 alloy. In particular, voids tended to preferentially aggregate on coarse Cu-rich and fine Mg2Si precipitates which contributes to the higher irradiation tolerance in terms of void formation for 6061 alloy, suggesting that the uniformly distributed precipitates can act as trapping sites for defect clusters during irradiation and thus promote recombination. The size and density of Mg2Si precipitates showed a slight increase and Cu-rich precipitates emerged in 6061 alloy after neutron irradiation. Moreover, radiation-induced segregation was also detected along the grain boundaries. After tensile testing, the pure aluminum showed a lower yield strength and larger uniform elongation than 6061 alloy, where a moderate uniform elongation was still reserved after irradiation to 4.2dpa. The severely reduced ductility in 6061 alloy is supposed to be caused by the transmutation product silicon which contributes to the increase of dispersed multiple precipitates after irradiation. Our results indicate that precipitates especially Mg2Si play a decisive role in void suppression, but are not suitable in diminishing the radiation hardening.

Multicentre investigation of neutron contamination at cardiac implantable electronic device (CIED) location due to high-energy photon beams using passive detectors and Monte Carlo simulations

Radiotherapy treatments involving LINACs operating at accelerating potentials >10 MV generate (photo)neutrons which deliver dose to patients also outside the target volume. This effect is particularly relevant for patients with cardiac implantable electronic devices (CIEDs), which can be damaged by the therapeutic irradiation. In the last few years, there has been a rising interest in this issue, and it seems that damage to CIEDs is primarily associated with the thermal component of the photoneutron field. In particular, a recent study led by Politecnico di Milano considered CIEDs from various manufacturers and showed that some of these devices can be damaged after an irradiation with a thermal neutron fluence of about 10^{9}hbox { cm}^{-2}. The present work results from a collaboration among Politecnico di Milano, the University of Pisa, the University of Trieste and three Italian hospitals located in Lucca, Trieste and Varese, respectively, and it is primarily aimed at evaluating the thermal neutron fluence in CIED region for some high-energy treatments delivered at 15 and 18 MV and to determine whether it is comparable to the critical value given above, which has been experimentally determined to be potentially harmful for CIEDs. Thermal neutron fluence was measured through CR-39 detectors and TLDs, which were housed inside a BOMAB-like phantom mimicking the patient’s trunk. The experimental sessions involved two models of LINAC, Varian Clinac DHX (Varese hospital) and Elekta Synergy (Lucca and Trieste hospitals). The experimental results show that the treatments considered in this study can lead to a thermal neutron fluence in the cardiac region comparable to the critical value. Furthermore, detailed Monte Carlo geometries for the facilities involved in this project were developed with the MCNP code (v. 6.2), and they were tested by comparing simulation results to measurements considering some benchmark irradiation plans. Bubble detectors were also employed for fast neutron fluence measurements to be compared to simulation outputs. These computational models stand out as promising tools for the investigations required in this work, and they can be used for further studies also extending their use to analogous facilities hosting the same models of LINACs.