Neutron spectrum measurements near KAMINI reactor south beam vault door using nested neutron spectrometer.

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275-300 µSv/h and 175-200 µSv/h, respectively, when the reactor is operating at 10 kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178 keV. Apart from energy reduction, 10 kW normalised neutron flux of south beam pit is also attenuated by the shield by 25 000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated.

Role of independent research at AERB for ensuring safety of nuclear facilities in India.

The mission of Atomic Energy Regulatory Board (AERB) of India is to ensure that the use of ionising radiation and nuclear energy in India does not cause unacceptable impact on the workers, members of the public and to the environment. AERB has the mandate to carry out detailed safety review for the siting, construction, commissioning, operation and decommissioning of nuclear and radiation facilities established within the country. To deliver and maintain a strong, credible and technically sound regulation, AERB has established the Safety Research Institute (SRI) at Kalpakkam with a robust technical infrastructure and wide knowledge base. This paper highlights the independent safety research activities carried out at SRI and its role to support and facilitate the decision-making process by AERB at various stages of regulatory review for ensuring safety of the nuclear facilities in India.

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.

An integrated method for the leakage fault mode diagnosis and life prediction of the reactor coolant pump.

The reactor coolant pump is a key equipment in a nuclear power plant. If the leakage exceeds a certain threshold, it may cause reactor overheating and shutdown. The reactor coolant pump leakage fault usually has two problems: corrosion and scaling. Accurately and efficiently diagnosing the leakage fault mode as early as possible and predicting its remaining useful life (RUL) are important for taking timely maintenance measures. In this paper, an integrated method is proposed. First, the cross-sectional area of the first seal is extracted as a fault indicator. The motivation is that corrosion may enlarge the cross-sectional area, and scaling may reduce the cross-sectional area. Based on the fluid mechanics theory, an integrated model with several uncertain parameters is established among the cross-sectional area, temperature, and leakage at the inlet and outlet of the first seal. In the diagnosing process, a modified change-detection method is proposed to detect the starting point of degradation. Then, the unknown parameters in the previous relation are estimated, and the degrading data before the starting point of degradation are used to diagnose the leakage fault mode. Second, a time-series model of the autoregressive integrated moving average (ARIMA) is established to predict the remaining useful life based on the degrading data after the starting point of degradation. Finally, the leakage degrading data from six reactor coolant pumps of a nuclear power plant is used to perform the leakage fault mode diagnosis and life prediction with degradation point detection error rates not exceeding 4%, fault mode diagnosis correction rates 100% and practical RUL predicting results, which proves that the proposed integrated method is accurate and efficient. The proposed integrated method combines the advantages of both the physical model diagnosis and the data-driven model diagnosis and innovatively make use of the quantity of flow from the output side of the primary pump as the monitoring indicator and the cross-sectional area as the characteristic index together to diagnose the leakage fault mode happened to the seal and predict its RUL, which can meet the needs of actual operation and maintenance to ensure a healthy and stable operation of the pump and prevent unexpected shutdowns of nuclear power plants and serious accidents.

Assessment of Occupational Radiation Exposures at Ghana Research Reactor-1 Facility.

The annual occupational doses for workers at the Ghana Research Reactor-1 facility were assessed for the period 2018-2021. The dose records of monitored staff were retrieved and analysis done for dose distribution and collective effective doses. Thermoluminiscent dosimeters were used to monitor the occupational exposures. The dosimeters were evaluated for the cumulative radiation dose levels using the Harshaw 6600 TLD reader system. Annual dose of 1.52 mSv/year was the maximum acquired by an individual. An annual average effective dose range of 0.20-1.36 mSv was determined for all workers. The annual total collective effective dose was established to be in the range of 0.40-10.08 man-Sv. The 20 mSv annual limit for occupational exposure was not exceeded for monitored workers. The assessment shows that the GHARR-1 facility, in terms of radiation health effects, is a favorable environment for workers since exposures are mostly below occupational exposure limit.

High-radiation-exposure tasks in Korean pressurized heavy-water reactors.

This study analysed the occupational dose in Korean pressurized heavy-water reactors (PHWRs) and identified tasks involving high radiation exposure. The average individual dose was sufficiently low to be below the annual effective dose limit for radiation workers and is even lower than the dose limit for the general public. However, some workers received relatively higher doses than others. Furthermore, most PHWR workers are exposed to radiation during planned maintenance periods. In this study, the radiation dose was normalized (radiation dose per unit time) to determine the high-radiation-exposure tasks in Korean PHWRs. Consequently, end-fitting lapping, delayed neutron tube work and fuel channel fixed-end change tasks were identified as high-radiation-exposure tasks in Korean PHWRs. If appropriate radiation protection measures are prioritized for the identified high-dose exposure tasks, optimization of radiological protection will be effectively achieved by reducing the dose that is relatively higher than the average.

Gross alpha/beta and radionuclide activity concentrations in soil, plant and some fruits around the Tehran Research Reactor.

Human activities usually have some contamination as effluents from chemical industries and radionuclides from nuclear reactors. For assessing the probable radioactive contamination in vicinity of Tehran Research Reactor, The gross alpha and beta radioactivity concentrations in soil, pine and cedar leaves and some selected fruits (fig, apple, berry and pomegranate) were investigated using an alpha/beta spectrometer during 2021-2022. Also, the concentrations of artificial and natural radionuclides in samples were investigated by the method of gamma spectroscopy. The gross alpha activity concentrations in soil, pine and cedar leaves and some selected fruits samples are from 0.05 to 0.35 Bq/gr and 0.07-0.31 Bq/gr and 0.04-0.18 Bq/gr, respectively. The gross beta activity concentrations in soil, pine and cedar leaves and some selected fruit samples are from 0.73 to 4.25 Bq/gr and 0.21-3.97 Bq/gr and 1.01-2.71 Bq/gr, respectively. Average activities concentration of natural radionuclide 232Th, 238U and 40K in soil, pine and cedar leaves and some selected fruits are 31.89-16.23-582.73 Bq/kg and 1.84-0.99-84.60 Bq/kg and 1.98-1.09-72.08 Bq/kg respectively. From artificial radionuclides, just 137Cs is recognized in soil sample and the range of 137Cs concentration in surface soils was observed to vary in the range 0.85-2.21 (Bq/kg). The result showed that the Tehran Research Reactor activities not have increased the environmental radioactivity and radiation level in the area.

Study of activation cross sections of double neutron capture reaction on 193Ir for the reactor production route of radiotherapeutic 195mPt.

The radiotherapeutic 195mPt is among the most effective Auger electron emitters of the currently studied radionuclides that have a potential theranostic application in nuclear medicine. Production of 195mPt through double neuron capture of enriched 193Ir followed by ß--decay to the radioisotope of interest carried out at the research reactor IBR-2 is described. Because of the high radiation background, radiochemical purification procedure of 195mPt from bulk of iridium was needed to be developed and is detailed here as well. For the first time, cross section and resonance integral for the reaction 194Ir(n,γ)195mIr were determined. Resonance neutrons contribution was established to exceed that of thermal neutrons, and resonance integral for the reaction 194Ir(n,γ)195mIr is calculated to be 2900 b. Specific activity of 195mPt was estimated to reach a value of 38.7 GBq/(g Pt) at IBR-2 by the end of bombardment (EOB).

Radiation source terms analysis and classification of radioactive waste for the decommissioning of the first HWRR reactor in China.

Nuclear reactors will face the problem of decommissioning at the end of their operating life due to the high radioactivity of reactor components and environmental safety considerations. The Heavy Water Research Reactor (HWRR) is the first large-scale research reactor to be decommissioned in China. The second phase of HWRR decommissioning involves the main components in the reactor block, so the radiation source terms and the radioactive waste level need to be evaluated before demolition and disposal. Based on the operating history, three-dimensional geometry, materials, and other information of the HWRR, the activity of radionuclides in the main components of HWRR is calculated and analyzed, and the MCNP/ORIGEN coupling scheme is utilized for theoretical analysis. The theoretical results indicate that 14C, 54Mn, 55Fe, 60Co, 63Ni, and 152Eu are the main radioactive nuclides. The total activity of radioactive nuclides was 2.36E + 15 Bq at the end of 2007, 4.27E + 13 Bq at the end of 2021, and 1.83E + 13 Bq at the end of 2025. Furthermore, local sampling and radiometric analyses based on the HPGe gamma-ray spectrometer are also performed to verify the theoretical results, the ratio of theoretical activity values to the measured activity of the experimental sample is within 2.5 times, so the theoretical results are conservative. According to the classification standards for radioactive waste, the inner shell, outer shell, cooling water tank, sand layer, and heavy concrete shielding layer are all low-level waste. These results and conclusions can serve as a reference for the second phase decommissioning of the HWRR and the subsequent disposal of radioactive waste.

Gamma-spectrometry measurements of filtration media from an Advanced Gas-cooled Reactor.

Filtration media used to quantify particulate and gaseous releases have been collected from Hartlepool Power Station in the United Kingdom and measured using high-sensitivity gamma-spectrometry systems. Radionuclides that are relevant to the monitoring regime of the Comprehensive Nuclear-Test-Ban Treaty (CTBT) have been detected. Results are reported and compared to detections recorded on the International Monitoring System (IMS). Time series activity plots have been produced and results interpreted with respect to known plant activities. The reported results improve the understanding of trace-level radionuclide emissions from Advanced Gas-cooled Reactors (AGRs) and aid interpretation of IMS measurements. This work is being performed as part of the Xenon Environmental Nuclide Analysis at Hartlepool (XENAH) collaboration between the Atomic Weapons Establishment (AWE, UK), EDF Energy (UK), Pacific Northwest National Laboratory (PNNL, US) and the Swedish Defence Agency (FOI, Sweden).

The colonization of an irradiated environment: the case of microbial biofilm in a nuclear reactor.

The investigation of the microbial community change in the biofilm, growing on the walls of a containment tank of TRIGA nuclear reactor revealed a thriving community in an oligotrophic and heavy-metal-laden environment, periodically exposed to high pulses of ionizing radiation (IR). We observed a vertical IR resistance/tolerance stratification of microbial genera, with higher resistance and less diversity closer to the reactor core. One of the isolated Bacillus strains survived 15 kGy of combined gamma and proton radiation, which was surprising. It appears that there is a succession of genera that colonizes or re-colonizes new or IR-sterilized surfaces, led by Bacilli and/or Actinobacteria, upon which a photoautotrophic and diazotrophic community is established within a fortnight. The temporal progression of the biofilm community was evaluated also as a proxy for microbial response to radiological contamination events. This indicated there is a need for better dose-response models that could describe microbial response to contamination events. Overall, TRIGA nuclear reactor offers a unique insight into IR microbiology and provides useful means to study relevant microbial dose-thresholds during and after radiological contamination.

The influence of some rare earth elements as neutron absorbers on the inception of Oklo natural nuclear reactors.

A totally reflected core model was built to estimate the infinite multiplication factor ${{k}}_{{\infty}}$ as a parametric function using MCNP code. Thus, it was possible to evaluate the influence of a specific physical parameter on the criticality occurrence of the Oklo phenomenon, namely initial Poisons (IP: Gd, Sm and Nd). In fact, these rare earth elements, prior to criticality occurrence in Oklo reaction zones (RZs), are considered as a key parameter in the present study. Thus, it was possible to construct isocritical lines, ${{k}}_{{\infty}}\left({{V}}_{{UO}{2}},{{\varPhi}}_{{C}}\ \right)\cong{1}$, over a significant range of Uraninite volume fraction: ${{V}}_{{UO}{2}}\left[\%\right]{\in}\left[{0};{40}\right]$. The corresponding critical porosity ${{\varPhi}}_{{C}}$ is obtained for a given value of ${{V}}_{{UO}{2}}$ by Python-driven MCNP5 calculations. By including realistic measurements of IP for different RZs, it was possible to distinguish the corresponding inception circumstances for the natural RZs considered here.

Detection limit variation against coarse aerosol during airborne contamination measurement with continuous air monitor.

In nuclear facilities, the mandatory airborne contamination surveillance is operated by continuous air monitors (CAMs). It samples the ambient air on a filter and measures the deposited activity. It is designed to trig an alarm whenever the measured activity concentration exceeds a defined threshold. However, in some sites, such as dismantling nuclear sites, a high rate of false alarm is experienced, mainly for artificial alpha. It has been shown that false alarms are directly related to a sudden variation of the aerosol mass sizes distribution, i.e. a wrong detection limit (DL) evaluation. Experiments on the ICARE tests bench have been carried out to compare the CAM's DL and an estimated DL as a function of the aerosol characteristics, to the measured one. This new estimation shows significant improvements over the previous one and highlights the need to consider aerosol characteristics for a correct airborne contamination measurement.

Oklo: historic and lessons learned.

Two outstanding phenomena have taken place in earlier geological era where Gabon is now located: the presence of natural nuclear reactors and the appearance of a very elaborated form of life for its age. Calculations performed to establish the sustained fission history of Oklo site are presented first. Second, possible correlations between these two anomalies are discussed. Could the presence of ionizing radiation be the cause of genetic mutations? Today's isotopic measurements allow us to improve our understanding of the irradiation suffered by organic matter over all times. The first objective is therefore to quantify the possible effects of such ionizing radiation. A second objective naturally appears: the storage of radioactive waste. Calculations issued from the first objective provide access to nuclear reactor waste formations and Oklo is the unique natural analogue of a long-term storage laboratory for nuclear waste. Returning to our primary objective, it is interesting to extend our reflections to other situations of naturally radioactive environments such as very old geological formations or lagoon.

Parallels between natural Oklo reaction zones and industrial reactors dynamics.

The first man-made nuclear reactor was developed by Fermi and collaborators at the University of Chicago and reached criticality in December 1942. This was the confirmation that men were able to use sustained fission reactions in order to produce energy. Following this success, nuclear reactors studies gave rise to several families of reactors corresponding to different orientations and technical choices. They are linked mainly to the choice of fuel (natural uranium, enriched uranium, plutonium, thorium), coolant (water, carbon dioxide, helium, sodium, ...) and moderator for slow neutron reactors (graphite, light water, heavy water). Out of all these choices, the pressurized water reactor (PWR) family is the closest to the Oklo natural reactors. Many intriguing similarities are observed and discussed in the present article. Our present-day understanding of the PWR operating conditions has been a great help for understanding the Oklo reactors.

United States' regulatory approved pharmacotherapies for nuclear reactor explosions and anthrax-associated bioterrorism.

INTRODUCTION: Nuclear reactor incidents and bioterrorism outbreaks are concerning public health disasters. Little is known about US Food and Drug Administration (FDA)-approved agents that can mitigate consequences of these events. We review FDA data supporting regulatory approvals of these agents. AREAS COVERED: We reviewed pharmaceutical products approved to treat Hematopoietic Acute Radiation Syndrome (H-ARS) and to treat or prevent pulmonary infections following Bacillus anthracis (anthrax) exposure. Four drugs were approved for H-ARS: granulocyte-colony stimulating factor (G-CSF), granulocyte/macrophage colony stimulating factor, pegylated G-CSF, and romiplostim. For bioterrorism-associated anthrax, the FDA approved five antibiotics (doxycycline, penicillin-G, levofloxacin, moxifloxacin, and ciprofloxacin), two monoclonal antibodies (obiltoxaximab and raxibacumab), one polyclonal antitoxin (Anthrax Immune Globulin Intravenous) and two vaccines (Anthrax Vaccine Adsorbed and Anthrax Vaccine Adsorbed with an adjuvant). A national stockpile system ensures that communities have ready access to these agents. Our literature search was based on data included in drugs@FDA (2001-2023). EXPERT OPINION: Two potential mass public health disasters are aerosolized anthrax dissemination and radiological incidents. Five agents authorized for anthrax emergencies only have FDA approval for this indication, five antibiotics have FDA approvals as antibiotics for common infections and for bacillus anthrax, and four agents have regulatory approvals for supportive care for cancer and for radiological incidents.

Progress with the regulation of radiation safety during recovery and removal of spent nuclear fuel from the site for temporary storage at Andreeva Bay on the Kola Peninsula.

In the 1960s, a shore technical base (STB) was established at Andreeva Bay on the Kola Peninsula, in northwest Russia. The STB maintained nuclear submarines and the nuclear icebreaker fleet, receiving and storing fresh and spent nuclear fuel (SNF) as well as solid and liquid radioactive waste (RW). It was subsequently re-designated as a site for temporary storage (STS) for SNF and RW. Over time, the SNF storage facilities partly lost their containment functions, leading to radioactive contamination of workshops and the site above permitted values. The technological and engineering infrastructure at the site was also significantly degraded as well as the condition of the stored SNF. At present, the STS Andreeva Bay facility is under decommissioning. This paper describes progress with the creation of safe working measures for workers involved in site remediation and SNF recovery operations, including the determination of safe shift times in high radiation areas, as part of overall optimization of safety. Results are presented for the successful application of these measures in the period 2019-2021, during which time significant SNF recovery and removal operations were completed without incident. Significant important experience has been gained to support safe removal of remaining SNF, including the most hazardous degraded fuel, as well as recovery of other higher level RW and decommissioning of the old storage buildings and structures.

Improved Measurement of the Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay.

Reactor neutrino experiments play a crucial role in advancing our knowledge of neutrinos. In this Letter, the evolution of the flux and spectrum as a function of the reactor isotopic content is reported in terms of the inverse-beta-decay yield at Daya Bay with 1958 days of data and improved systematic uncertainties. These measurements are compared with two signature model predictions: the Huber-Mueller model based on the conversion method and the SM2018 model based on the summation method. The measured average flux and spectrum, as well as the flux evolution with the ^{239}Pu isotopic fraction, are inconsistent with the predictions of the Huber-Mueller model. In contrast, the SM2018 model is shown to agree with the average flux and its evolution but fails to describe the energy spectrum. Altering the predicted inverse-beta-decay spectrum from ^{239}Pu fission does not improve the agreement with the measurement for either model. The models can be brought into better agreement with the measurements if either the predicted spectrum due to ^{235}U fission is changed or the predicted ^{235}U, ^{238}U, ^{239}Pu, and ^{241}Pu spectra are changed in equal measure.

Inspect: a decision helping tool for in-situ probe selection in D&D activities.

In the context of the INSIDER European project, the suitability of existing methodologies for in-situ measurements under constraint environments in nuclear facilities following a decommissioning and dismantling (D&D) process was analysed. Firstly, an analysis of the different methodologies for in-situ measurements was made along with a study of the different types of constrained environments that could appear in the D&D process as well as their expected level of impact on the measurement methodologies. Based on this analysis, a decision-helping tool for the selection of the suitable in-situ equipment/detector to be used in nuclear facilities for the different phases in any D&D process has been developed, depending on the constrained environment. This tool is named INSPECT, acronym for In-Situ Probe SelECtion Tool. The software is therefore potentially of use to those working in radiological characterization with in-situ instrumentation in any radiological or nuclear D&D process.

Production and characterization of 111Ag radioisotope for medical use in a TRIGA Mark II nuclear research reactor.

Radio Pharmaceutical Therapy (RPT) comes forth as a promising technique to treat a wide range of tumors while ensuring low collateral damage to nearby healthy tissues. This kind of cancer therapy exploits the radiation following the decay of a specific radionuclide to deliver a lethal dose to tumor tissues. In the framework of the ISOLPHARM project of INFN, 111Ag was recently proposed as a promising core of a therapeutic radiopharmaceutical. In this paper, the production of 111Ag via neutron activation of 110Pd-enriched samples inside a TRIGA Mark II nuclear research reactor is studied. The radioisotope production is modeled using two different Monte Carlo codes (MCNPX and PHITS) and a stand-alone inventory calculation code FISPACT-II, with different cross section data libraries. The whole process is simulated starting from an MCNP6-based reactor model producing the neutron spectrum and flux in the selected irradiation facility. Moreover, a cost-effective, robust and easy-to-use spectroscopic system, based on a Lanthanum Bromo-Chloride (LBC) inorganic scintillator, is designed and characterized, with the aim of using it, in the future, for the quality control of the ISOLPHARM irradiated targets at the SPES facility of the Legnaro National Laboratories of INFN. natPd and 110Pd-enriched samples are irradiated in the reactor main irradiation facility and spectroscopically characterized using the LBC-based setup and a multiple-fit analysis procedure. Experimental results are compared with theoretical predictions of the developed models, showing that inaccuracies in the available cross section libraries prevent an accurate reproduction of the generated radioisotope activities. Nevertheless, models are normalized to our experimental data allowing for a reliable planning of the 111Ag production in a TRIGA Mark II reactor.