Reactor Facility Research Articles

Characterization of the PGNAA neutron beam of Isfahan MNSR through the calculations and different measurement methods

In this work, the significant characterization of the neutron beam used in the PGNAA facility of the Isfahan MNSR research reactor has been determined using both calculation and measurement. The important parameters of the beam include flux and energy distribution of neutrons, cadmium ratio, thermal neutron content (TNC), and neutron-to-gamma ratio (n/γ). In this regard, Monte Carlo simulation and experiments were implemented to find these characteristics. The experiments have been carried out through the relative and absolute activation methods using irradiation of indium foil and mica in combination with uranium foil. Moreover, the condition of the neutron beam has been investigated in terms of contamination with fast neutrons using the germanium triangle method. The results showed that the MNSR has acceptable characteristics for performing PGNAA analyses despite its low power.

An experimental and modeling study of propane oxidation kinetics in low temperature supercritical water

Propane oxidation in supercritical water was investigated at iso-thermal iso-baric conditions using a batch reactor facility. Mixtures were comprised of 0.014 % propane by volume with an equivalence ratio of 0.8 and a total density of 222 mg/mL or 610 mg/mL. Reaction times ranged from 8 to 30 min for a temperature of 375ºC at 220 or 400 bar, or 400ºC at 220 bar. Major reaction products were CO and CO2 and minor products were propene, acetone, ethene, ethanol, methane, methanol and hydrogen. New detailed chemical kinetic models were developed by combining and refining existing models using genetic optimization. Model predictions exhibited excellent agreement with experimental observations, and indicated that rates of H-abstraction and OH addition reactions involving alkanes and alkenes are affected by the supercritical water environment. Model accuracy was highly sensitive to the rates of CH3O2H = CH3O + OH and CH3 + H2O2 = CH4 + HO2.

COOLING OF CONTROL RODS AND SPECIFICATION OF CRITICAL TEMPERATURES TO JUSTIFY SAFE OPERATION

The paper analyzes the cooling conditions of the control rods in the mixed core of the WWER-1000 reactor containing fuel assemblies of different vendors (TVSA, WFA/RWFA). The coolant heat-up in the guide thimbles due to own energy release in control rod materials and control rod temperatures required to justify safe operation were calculated for the most conservative case, when the core consists only of TVSAs with a standardized guide thimble inlet plug of NCCP. The energy release was obtained using the calculation method for the boron carbide absorber (B4C), dysprosium titanate (Dy2O3·TiO2), control rod cladding (42CrNiMo), water gap between the cladding and the guide thimble and in the guide thimble (zirconium alloy), fully immersed and 70% immersed RCCA 10th control group. The highest calculated energy release will be observed in fully immersed control rods with boron carbide material at the height position corresponding to the zone of maximum neutron fluence and will be 43.8 W/cm3. The maximum energy release of the 70% immersed (from the bottom of the core) 10th control group is 29.4 W/cm3 for fresh B4C. For the condition of the reactor facility at which the temperatures in FA guide thimbles are maximum, for a "fresh" control rod characterized by maximum energy release in the neutron-absorbing material, the maximum coolant heat-up in the guide thimbles does not exceed 28.9 °C, with the maximum coolant heat-up in FAs being 44 °C, and the maximum control rod cladding temperature not exceeding 335.6 °C at a coolant saturation temperature of 345.8 °C, which indicates that there will be no coolant boiling in the guide thimbles under operation. The maximum temperature typical for the central part of the control rod absorber and for the most conservative calculation for a "fresh" absorber is 529.6 °C, which is significantly lower the melting point of boron carbide, which is ~2350 °C. The maximum calculated temperature in the contact zone of the B4C absorber with control rod cladding is 348 °C. The performed calculations and further analysis of the results confirm that during operation of the RCCA control rods with both “fresh” and “burned-up” B4C, there will be no boiling of the coolant and melting of the absorber in the central part of the control rod. This means that the VVER-1000 core with TVSAs and WFAs/RWFAs and the RCCA rods are reliably cooled.

Artificial radionuclides in the plant cover around nuclear fuel cycle facilities.

This paper presents research on the assessment of the radioecological state of plant cover surrounding two research reactor facilities located within the Semipalatinsk Test Site (STS) as examples of nuclear fuel cycle facilities (NFC). Source data on the concentrations of artificial radionuclides in the plant cover were obtained. Quantitative values for 137Cs, 241Am, and 239+240Pu activity concentrations were determined in plants across the perimeters of the facilities, indicating that these compounds may be present in the associated media from the perspective of accumulative bioindication. The values determined for artificial radionuclides in the 'soil‒plant' system around the researched NFC facilities were attributed to radioactive contamination of the STS territory.

Present and new operational quantities evaluated from photon spectrum measurements at workplaces in the research reactor and accelerator facility at the JAEA

The International Commission on Radiation Units and Measurements (ICRU) has proposed to change the definitions of the operational quantities used for the area and individual monitoring of external exposure in ICRU Report 95. Since introducing new operational quantities into the radiation monitoring may affect the dose assessment using the present personal dosimeters, precise assessments of the influence on monitoring are necessary by characterizing the energy spectrum in the workplace and the energy dependency of the dosimeters to be used. This study measured photon spectra using a NaI(Tl) scintillation detector or a LaBr3(Ce) scintillation detector at the Japan Research Reactor No. 3 (JRR-3) and Japan Proton Accelerator Research Complex (J-PARC) workplaces at the Japan Atomic Energy Agency (JAEA). The photon energy spectra were obtained by unfolding the pulse height distributions measured at each workplace. The present and new operational quantities were then evaluated and compared using the spectra measured at the workplaces.In JRR-3, the photon energy spectra were measured at the workplaces in the reactor room and thermal neutron beam hall while the reactor was operated. The ambient dose equivalent rates H˙*(10) at the workplaces were 0.2 μSv·h−1 – 13 μSv·h−1. High energy gamma-rays of 6–7 MeV from 16N were observed near the fuel failure detection system in the reactor room. Those of 7–8 MeV produced by the thermal neutron capture of iron were observed in the reactor room and the beam hall. In J-PARC, the photon energy spectra were measured at the workplaces in the beam tunnel of the 3 GeV synchrotron beam ring five days after the operation. The ambient dose equivalent rates H˙*(10) at the workplaces were 35 μSv·h−1 –430 μSv·h−1.It was found that the new personal doses, Hp, were 10%–20% smaller than the present personal dose equivalents, Hp(10), at the workplaces.

Thermodynamic Analysis and Optimization of Binary CO2-Organic Rankine Power Cycles for Small Modular Reactors

Small nuclear power plants are a promising direction of research for the development of carbon-free energy in isolated power systems and in remote regions with undeveloped infrastructure. Improving the efficiency of power units integrated with small modular reactors will improve the prospects for the commercialization of such projects. Power cycles based on supercritical carbon dioxide are an effective solution for nuclear power plants that use reactor facilities with an initial coolant temperature above 550 °C. However, the presence of low temperature rejected heat sources in closed Bryton cycles indicates a potential for energy saving. This paper presents a comprehensive thermodynamic analysis of the integration of an additional low-temperature organic Rankine cycle for heat recovery to supercritical carbon dioxide cycles. A scheme for sequential heat recovery from several sources in S-CO2 cycles is proposed. It was found that the use of R134a improved the power of the low-temperature circuit. It was revealed that in the S-CO2 Brayton cycle with a recuperator, the ORC add-on increased the net efficiency by an average of 2.98%, and in the recompression cycle by 1.7–2.2%. With sequential heat recovery in the recuperative cycle from the intercooling of the compressor and the main cooler, the increase in efficiency from the ORC superstructure will be 1.8%.

Assessment of radiological consequences to a hypothetical accident of the 3-MW TRIGA Mark-II nuclear research reactor of Bangladesh

The radiological consequences to a hypothetical accident are assessed for the 3-MW TRIGA Mark-II Nuclear Research Reactor of Bangladesh for both wet and dry seasons. Accident simulations are crucial for assessing the potentialrisks in the event of a nuclear accident. ORIGEN code was used to carry out the facility's radionuclide inventory. In the event of a hypothetical accident, the Total Effective Dose Equivalent (TEDE) is assessed using HotSpots. The findings indicate that for stability class A during the rainy season, the selected radionuclides arrived at a downwind distance of 0.03 km with an arrival time of less than one minute, and the highest TEDE, respiratory time-integrated air concentration, and ground deposition activity, respectively, were marked at 2.8E-12 Sv, 9.4E-01 Bq-sec/m3, and 2.4E-05 kBq/m2. In the case of I-131 at 0.03 km from the source, the thyroid gland is the most sensitive human organ, followed by the lung, esophagus, thymus, surface bone, skin, and breast. The estimated Committed Effective Dose Equivalent (CEDE) from I-131 is 9.1E-11 Sv. In contrast, the adrenal was the most sensitive organ with a maximum CEDE of 1.3E-13 Sv in the case of Cs-137 at a downwind distance of 0.03 km. The highest TEDE values for the general public and the vicinity around the TRIGA Mark II reactor facility are lower than the annual dose limits.

Capture enhanced neutron irradiation for the treatment of the Alzheimer Disease: Micro-nanodosimetric characterization of beta amyloid peptide samples

Alzheimer Disease (AD) is the most common cause of dementia. 46.8 million people live with AD worldwide, with numbers projected to almost double every 20 years. The etiological mechanisms underlying the neuropathological changes in AD remain unclear. The beta amyloid peptide Aβ is considered to be the main culprit of the pathological processes. NECTAR (NEutron Capture-enhanced Treatment of neurotoxic Amyloid aggRegates) project proposes an alternative and revolutionary strategy to address AD, investigating the safety, feasibility and effectiveness of a Capture-Enhanced Neutron Irradiation to structurally damage Aβ aggregates, by exploiting capture agent vectors containing B-10 and Gd-157.Among many others, one of the goals of the project is to perform calculations and measurements of microdosimetric and nanodosimetric quantities to characterize the effect of reaction products at local level and to link it to the fibrils depolymerization process. In this framework, a new low-pressure tissue-equivalent proportional counter (TEPC) was designed and developed. The system is a wall-free TEPC able to measure micro-nanometric dose deposition for different biological samples. It is characterized by two main features: i) a wall-less structure allows the charged particles produced by the sample to access the sensitive volume, where their effects are studied; ii) a trigger system enables the assessment of the number of ionizations induced in the sensitive volume related to a single primary particle for obtaining nandosimetric information.The prototype system was tested in the PGNAA (Prompt Gamma Neutron Activation Analysis) facility of the LENA reactor (Pavia) to directly study secondary particles from Aβ fibrils, loaded with Boron. Results of the micro-nanodosimetric characterization of high LET secondaries from Aβ-fibrils were compared to a reference Boron sample. For a 500 nm simulated site size, the presence of Boron in the Aβ sample greatly populates microdosimetric spectra at high lineal energy values. Instead, lower LET particles (photons, protons) from reactions on the other Aβ aggregate solution molecules populate the low lineal energy portion of the distribution.The nanodosimetric distributions highlights the presence of different field components on the nanodosimetric spectra: low LET particles (protons/electrons) produce small ionization clusters, while alpha and Lithium ions induce larger clusters, thus have a higher biological effect. The nanodosimetric measurements have been compared with Monte Carlo simulations performed with a track-structure code, showing a good agreement.This multi-level approach provides a deeper physical understanding of the mixed secondaries field coming from B-enriched target, as well as a reference for the biological effect on Aβ fibrils at nanometer level. These preliminary measurements represent the first microdosimetric and nanodosimetric characterization of the secondary mixed field directly emitted by a biological sample irradiated with neutrons.

Passive single-sphere multi-detector spectrometry system for profiling of neutron fields in the thermal to 20 MeV energy range

This study explores the development and application of a custom-designed single-sphere multi-detector spectrometer (SSMDSS) for characterizing neutron fields. Stray neutrons in reactor and accelerator facilities present complex, energy-diverse fields. Accurately assessing these fields is crucial for personnel safety. The SSMDSS system is designed to address this challenge, offering a compact and passive dosimetry solution for constrained workspaces. We studied the performance of SSMDSS in two environments: radio-nuclide based neutron sources and stray neutron fields at a research reactor hall. Our investigation demonstrates that SSMDSS provide reliable neutron energy distributions. The performance of the SSMDSS was compared with commercially available ROSPEC system. SSMDSS exhibits superior resolution due to its design, leading to a more accurate measurement of low-energy fluence when compared with ROSPEC. This distinction is reflected in various measured parameters, including average energy and fluence-to-dose conversion coefficients. Inside the reactor hall, SSMDSS measured a fluence-to-dose conversion coefficient of 121 pSv.cm2 compared to ROSPEC's 83 pSv.cm2. Additionally, SSMDSS measured a higher average energy (0.25 MeV) compared to ROSPEC (0.14 MeV). While ROSPEC remains a valuable tool for real-time data collection in dynamic fields with high gamma backgrounds, SSMDSS presents itself as a viable alternative for neutron spectrometry in neutron work fields due to its compact design and passive dosimetry capabilities. This makes it particularly suitable for deployment in constrained workspaces with potential stray neutron exposure.

Prototyping a High Purity Germanium cryogenic veto system for a bolometric detection experiment

The use of High Purity Germanium detectors operated in ionization mode at cryogenic temperatures is investigated as an external background mitigation solution for bolometers used in rare-event search experiments. A simple experimental setup, running a 52-g Li2WO4 bolometer sandwiched in-between two 2-cm thick High Purity Germanium cylindrical detectors in a dry cryostat, shows promising rejection to environmental gammas and atmospheric muons backgrounds. The acquired data are used together with a Monte Carlo simulation of the setup to extract the main contributions to the external backgrounds expected in an above ground experiment, such as e.g. current and future experimental efforts targeting the detection of coherent elastic neutrino-nucleus scattering at reactor facilities. Based on all these results, a 4π coverage similar veto system achieving a O(10 keV) energy threshold is expected to achieve a ≳73% and a ≳92% rejection power for gamma-like and muon-like events, respectively.

ANALYSIS OF METHODS FOR SIMULATING THE DECAY HEAT IN CORIUM WHEN MODELING A SEVERE ACCIDENTS AT NUCLEAR POWER PLANT

It is known that during development of a severe accident at a nuclear power plant, the melting of core materials and theformation of corium occurs. A feature of corium is the presence of a decay heat, which contributes a lot to the nature of its interaction with the structural materials of the reactor facility. In this regard, quite serious requirements are imposed on methods for simulating decay heat in the corium prototype, which relate to both the uniformity of the volume distribution and its intensity. This paper presents a comparative analysis of existing methods for decay heat simulation in corium, which are used at various experimental facilities investigating the operation of passive protection systems in severe accidents with reactor meltdown at nuclear power plants. By comparing the advantages and disadvantages, a more practical method of decay heat simulation is determined and ways are proposed to further improve the chosen method to fully simulate the thermal field of a real corium.

Optimization and Applications of Instrumental and Preconcentration Neutron Activation Analysis Methods for Rapid Determination of Selected REE by Utilizing their Short‐Lived Activation Products**

AbstractDetermination of Rare Earth Elements (REEs) using spectroscopic methods is challenging. Conventional Instrumental Neutron Activation Analysis (INAA) method utilizing Pneumatic Carrier Facility (PCF) of Dhruva research reactor has been optimized with the short‐lived isotopes such as 161Gd, 165mDy and 153/155Sm for quantification of Gd, Dy and Sm respectively. Such isotopes are difficult to observe by conventional INAA with longer cooling periods owing to their very short half‐lives (1–3 mins). In this work, such isotopes are not only observed but also utilized as activation product for rapid determination of these REEs. The standardized method has been utilized for the first time in liquid samples involving nuclear fuel like U3O8 wherein TBP/n‐dodecane has been applied for chemical separation of Uranium prior to analysis. The INAA method has also been applied for validating the stoichiometry of Gd in Gd3+ co‐doped Zn2P2O7:Tb3+ green emitting solid‐state phosphors, which plays a very pivotal role in sensitizing the terbium ion in its performance for solid state lighting. We believe with huge current demand of REEs in research, science and technology; INAA can be a game changer wherever needed for quantification with high precision and accuracy and can also be extended to other REEs of interest.

Fluence profiling at JSI TRIGA reactor irradiation facility

We present an analysis of the fluence profile at the JSI TRIGA neutron reactor facility in Ljubljana. For the study, multi-pad Low-Gain Avalanche Diodes (LGADs) are used. The deactivation of acceptor doping in the gain layer implant due to the irradiation, typical of LGAD devices, is exploited to map the fluence profile inside the irradiation channels. The amount of active doping of the LGAD gain layer is extracted via capacitance–voltage measurements for each pad before and after irradiation to a fluence of 1.5 × 1015neq/cm2, where neq stands for 1 MeV equivalent neutron count, providing a precise and prompt measurement of the fluence distribution over the LGAD sensor. Experimental results are compared to neutron fluence expectations calculated with Monte Carlo techniques.

A Consequence-Informed Licensing Path Selection for the Design of Physical Protection Systems at Commercial Nuclear Power Facilities

The survivability of the domestic nuclear power industry depends on the cost-competitiveness of safe and secure nuclear power generation. Advanced reactor design concepts aim to have increased safety margins over traditional large light water reactors (LWRs). With increased safety margins comes the potential for a corresponding decrease in off-site risk to the general public from a hypothetical release of radioactivity due to sabotage or theft. Without sacrificing safety or security, advanced reactor designers may be able to achieve operational cost improvements over current LWRs in part by designing less burdensome physical protection systems (PPSs) and by replacing on-site response forces with off-site response forces. To accommodate these developments, the U.S. Nuclear Regulatory Commission is drafting new rulemakings for physical security when licensing through the current frameworks in 10 CFR 50 or 10 CFR 52 along with drafting an entirely new licensing framework: 10 CFR 53. A novel technology-inclusive consequence-informed methodology for the selection of the optimal licensing path for the design of PPSs at advanced fixed-site commercial nuclear power facilities is presented herein. This methodology proposes integrating security considerations at the beginning of a reactor facility design effort to streamline the licensing process. Off-site total effective dose equivalents at the exclusion area and low population zone boundaries were identified as the key metrics when determining a design’s most appropriate licensing path that in turn affects the design requirements placed upon the PPS. Given these metrics, source-term generation of potential adversary-induced physics-based sabotage actions utilizing severe accident modeling software and off-site plume dispersal modeling were identified as appropriate for determining siting constraints, potential target sets for hypothetical sabotage events, and their subsequent off-site dose consequences. The methodology proposes using the consequence results from the sabotage modeling, in combination with desired cost-saving PPS characteristics, to help inform the licensing path selection. Once a licensing path is chosen, the methodology utilizes the Design and Evaluation Process Outline to evaluate an effective PPS following the licensing requirements placed on the facility. This paper also presents examples of hypothetical commercial nuclear power facilities with varying consequence levels and demonstrations of how to select the optimal licensing pathways for each.

An extension of NASCA for SDDR analysis of mobile activated components with global variance reduction

During the plasma operation period of fusion reactor facilities, neutrons generated by DT reaction can penetrate deeply into structures and induce strong activation. Highly activated components have to be transferred from the on-load position to the hot cell for the purpose of maintenance. This movement produces a high-dose field, originating from the mobile radiation source, which could affect the functionality of important devices or pose risks to the workers. Therefore, it is imperative to compute the shutdown dose rate (SDDR) maps while transporting activated components in the facility and develop an optimized solution for this calculation process. In this study, we established a new extension of the NASCA method combined with the OTF GVR method to compute SDDR maps. NASCA-OTF is a method developed based on the shutdown dose simulation code NASCA to ensure automatic modification of the geometry structure on the fly. In the meantime, it uses OTF method for speeding up the Calculation. Thus, this modified method produces convergent SDDR maps corresponding to different positions of activated components within a single Monte Carlo (MC) simulation. The NASCA-OTF was verified by comparison with R2S method through benchmark analysis. The method was also used for SDDR analysis of the mobile blanket sector for the China Fusion Engineering Test Reactor (CFETR), and the results indicated that the NASCA-OTF is feasible for computing the SDDR resulting from the movement of activated components.

Cooling System of Research Reactor Facility as Basis for Boosted Regression

Cooling System of Research Reactor Facility as Basis for Boosted Regression

Pre-Experimental Assessment of Uranyl Nitrate Solution Irradiation in Kartini Reactor Facilities

Medical radioisotope production using neutron irradiation via fission reaction requires a sufficient neutron source. The Kartini reactor has been proposed and studied to become a neutron source for radioisotope production under the Subcritical Assembly for 99Mo Production (SAMOP) project, which uses uranyl nitrate solution as the irradiation target. A full-scale experiment involving a liquid fission product is difficult to conduct and requires facility rearrangement to reduce the risk of contamination. Although a small-batch experiment is safer to perform, a pre-experimental assessment is necessary to address the practicality of production and the accompanying problems. The goals of this assessment are (1) to characterize the Kartini reactor irradiation facilities’ flux through experiment and Monte Carlo benchmark simulation, (2) to predict the irradiation product inventory in relation to the variation of uranium concentration and the measured flux, and (3) to predict the irradiated sample gamma spectrum reading using high-purity germanium detector simulation. The irradiation simulation uses natural uranium as a control parameter, which caused the irradiation inventory dominated by actinides from transmutation. The simulation also presents the possibility of instant small-batch 99Mo production using the measured Lazy Susan facilities’ flux from a neutronic perspective. The qualitative assessment of the predicted irradiation inventory and its spectrum reading from different sample concentrations are discussed along with the recommendation and possible action to improve the experiment or future production process.

Gamma-ray Spectroscopy in Low-Power Nuclear Research Reactors

Gamma-ray spectroscopy is an effective technique for radioactive material characterization, routine inventory verification, nuclear safeguards, health physics, and source search scenarios. Gamma-ray spectrometers typically cannot be operated in the immediate vicinity of nuclear reactors due to their high flux fields and their resulting inability to resolve individual pulses. Low-power reactor facilities offer the possibility to study reactor gamma-ray fields, a domain of experiments hitherto poorly explored. In this work, we present gamma-ray spectroscopy experiments performed with various detectors in two reactors: The EPFL zero-power research reactor CROCUS, and the neutron beam facility at the Ohio State University Research Reactor (OSURR). We employed inorganic scintillators (CeBr3), organic scintillators (trans-stilbene and organic glass), and high-purity germanium semiconductors (HPGe) to cover a range of typical—and new—instruments used in gamma-ray spectroscopy. The aim of this study is to provide a guideline for reactor users regarding detector performance, observed responses, and therefore available information in the reactor photon fields up to 2 MeV. The results indicate several future prospects, such as the online (at criticality) monitoring of fission products (like Xe, I, and La), dual-particle sensitive experiments, and code validation opportunities.

Methods to maximize detector count rates on small-angle neutron scattering diffractometers at reactor sources: I. Optimizing wavelength selection

Methods to determine the optimal neutron wavelength to maximize detector count rates on small angle neutron scattering (SANS) diffractometers at reactor sources is presented. Three experimental methods are used to determine the choice of optimal wavelength and collimation combination that maximizes the detector count rate. The wavelength optimization methods are applied to two different SANS diffractometers at the NCNR, and all methods are found to confirm the optimal wavelength of approximately 9.5 Å ± 0.5 Å for optically thin samples. The optimum wavelength is shifted by the wavelength-dependence of the sample transmission to 8.5 Å ± 0.5 Å for thicker absorbing samples or 6.5 Å ± 0.5 Å if the amount of multiple scattering needs to be minimized to limit scattering curve distortions. The same optimization methods can be applied to SANS diffractometers at other reactor facilities.

Радиоэкологическая оценка реакторных установок на основе сценариев проектных и запроектных аварий

The methodical approach based on scenarios of design and beyond design basis accidents was developed for comparative radioecological assessment of reactor facilities. The dynamics of possible radiation doses to the thyroid and the whole body of the public resided nearby the Russian nuclear power plants equipped with the VVER-440, VVER-1000, VVER-1200 and BN-800 reactors was estimated. With the use of the conservative approach for each accident scenario, as well as with the use of acute and medium-term dose criteria, radiation exposure indices were calculated for each accident scenario. Reactor facilities are ranked according to an index summarizing the accident scenarios features and the probability of their realization, the height of radioactive emissions and the activity of dosage-forming radionuclides. The potential radioecological danger of Russian reactors to the population has been reduced. The lowest estimates of the summarized emergency risk are typical for the advanced VVER-1200 reactor of the 3+ generation and the fast neutron reactor BN-800. The most severe emergency scenarios for each considered reactor are highlighted. The VVER-440 reactor has the highest number of severe projected accidents exceeding the dosage criteria envisaging evacuation of the population. The directions of further research related to the analysis of the environmental radiological effects, the justification of radiation protection measures and the development of operational intervention levels in the acute post-accident period are outlined.