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This study investigates the feasibility of siting a nuclear power plant (NPP) within a densely populated university campus. Site selection for NPPs is a complex process involving numerous interconnected factors. To address this complexity, we employ system dynamics (SD) modeling, a methodology renowned for its ability to quantify both technological and social aspects. Our simulation results indicate that public acceptance (PA) is a critical determinant in site selection. In the fourth month of the simulation, the model's values were identical in both PA and non-PA scenarios. However, subsequent simulations revealed a significant divergence, with PA-implicated values reaching 85.398 compared to 24.240 in the non-PA scenario. This disparity underscores the primacy of PA in the decision-making process. Given the extended timeline of NPP construction, the gap between PA and non-PA outcomes is likely to widen further, particularly in the context of positive public sentiment toward the NPP. The insights gleaned from this research can be applied to other industries, such as oil refineries, where the potential for atmospheric pollution due to toxic gas leaks necessitates careful site selection. • ●The site selection of NPP is analyzed in a campus town. • ●PA is a factor to consider in the site decision. • ●Simulations show the quantified results. • Future site section could be applied using this method. [ABSTRACT FROM AUTHOR]

The National Criticality Experiments Research Center (NCERC) located at the Nevada National Security Site (NNSS) in the Device Assembly Facility (DAF) and operated by Los Alamos National Laboratory (LANL) is the only general purpose critical experiments facility in the United States. Experiments from subcritical to critical and above prompt critical are carried out at NCERC on a regular basis. In recent years, NCERC has become more involved in experiments related to nuclear energy, including the Kilopower/KRUSTY demonstration and the recent Hypatia experiment. Multiple nuclear energy related projects are currently ongoing at NCERC. This paper discusses NCERC’s role in advanced reactor design and how that role may change in the future.

A new measurement of the reactor antineutrino flux and energy spectrum by the Daya Bay reactor neutrino experiment is reported. The antineutrinos were generated by six 2.9 GWth nuclear reactors and detected by eight antineutrino detectors deployed in two near (560 m and 600 m flux-weighted baselines) and one far (1640 m flux-weighted baseline) underground experimental halls. With 621 days of data, more than 1.2 million inverse beta decay (IBD) candidates were detected. The IBD yield in the eight detectors was measured, and the ratio of measured to predicted flux was found to be 0.946±0.020 (0.992±0.021) for the Huber+Mueller (ILL+Vogel) model. A 2.9σ deviation was found in the measured IBD positron energy spectrum compared to the predictions. In particular, an excess of events in the region of 4-6 MeV was found in the measured spectrum, with a local significance of 4.4σ. A reactor antineutrino spectrum weighted by the IBD cross section is extracted for model-independent predictions.

Hydrogen is increasingly being viewed as a significant fuel for future industrial processes as it offers pathways to zero emission. The UK sees hydrogen as one of a handful of low-carbon solutions for transition to net zero. Currently, most hydrogen production is from steam reforming of natural gas or coal gasification, both of which involve the release of carbon dioxide. Hydrogen production from mini decentralised grids via a thermochemical process, coupled with electricity production, could offer favourable economics for small modular reactors (SMRs), whereby demand or grid management as a solution would include redirecting the power for hydrogen production when electricity demand is low. It also offers a clean-energy alternative to the aforementioned means. SMRs could offer favourable economics due to their flexible power system as part of the dual-output function. This study objective is to investigate the critical performance parameters associated with the nuclear power plant (NPP), the cycle working fluids, and control-system design for switching between electricity and hydrogen demand to support delivery as part of a mini grid system for a reactor power delivering up to approximately 600 MWth power. The novelty of the work is in the holistic parametric analysis undertaken using a novel in-house tool, which analyses the NPP using different working fluids, with a control function bolt-on at the offtake for hydrogen production. The results indicate that the flow conditions at the offtake can be maintained. The choice of working fluids affects the pressure component. However, the recuperator and heat-exchanger effectiveness are considered as efficiency-limiting factors for hydrogen production and electricity generation. As such, the benefit of high-technology heat exchangers cannot be underestimated. This is also true when deciding on the thermochemical process to bolt onto the plant. The temperature of the gas at the end of the pipeline should also be considered to ensure that the minimum temperature-requirement status for hydrogen production is met. [ABSTRACT FROM AUTHOR]

The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin22θ13 and the effective mass splitting Δmee2. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors' baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This paper describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

Abstract In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented.

Abstract In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented.

While many states are moving away from nuclear power and decommissioning their reactors for cheaper and safer alternative energy sources, recent years have seen a spark in interest for nuclear power within the Middle East under the pretext of ‘energy independence’. This trend poses a potential threat for the safety of the region considering that nuclear power plants are prone to human errors, deliberate attacks, and natural environmental convulsions which could trigger potential transboundary fallout. Given the region’s small and compacted geography along with the increasingly volatile geopolitical instability, a regional incident would likely have much direr consequences compared to other previous nuclear incidents. As a non-nuclear power state, Qatar is not currently party to any of the nuclear civil liability conventions which could guarantee some level of compensation for the victims in case of transboundary nuclear harm. In due course, Qatar will be surrounded by nuclear reactors from the north (the Iranian Bushehr plant), the east (the UAE Barakah plant) and the west (the planned Saudi plants). As exemplified by the COVID-19 pandemic, theoretical transboundary calamities can unexpectedly become a sudden reality and there is a solemn need to work proactively when dealing with such consequential hypotheticals. Therefore, this article qualitatively assesses the relevant international conventions with an aim of being policy relevant and navigate Qatari decision-makers through the vexing web of the nuclear civil liability regimes.

While many states are moving away from nuclear power and decommissioning their reactors for cheaper and safer alternative energy sources, recent years have seen a spark in interest for nuclear power within the Middle East under the pretext of ‘energy independence’. This trend poses a potential threat for the safety of the region considering that nuclear power plants are prone to human errors, deliberate attacks, and natural environmental convulsions which could trigger potential transboundary fallout. Given the region’s small and compacted geography along with the increasingly volatile geopolitical instability, a regional incident would likely have much direr consequences compared to other previous nuclear incidents. As a non-nuclear power state, Qatar is not currently party to any of the nuclear civil liability conventions which could guarantee some level of compensation for the victims in case of transboundary nuclear harm. In due course, Qatar will be surrounded by nuclear reactors from the north (the Iranian Bushehr plant), the east (the UAE Barakah plant) and the west (the planned Saudi plants). As exemplified by the COVID-19 pandemic, theoretical transboundary calamities can unexpectedly become a sudden reality and there is a solemn need to work proactively when dealing with such consequential hypotheticals. Therefore, this article qualitatively assesses the relevant international conventions with an aim of being policy relevant and navigate Qatari decision-makers through the vexing web of the nuclear civil liability regimes.

Among a series of important safety parameters necessary for a complete analysis of a nuclear reactor core, the nuclear safety operating margin of a nuclear reactor is of great importance. The methodology generally used is to combine thermalhydraulic calculations with neutronic calculations. In this work, new Power Peaking Factors (PPFs) are calculated for the NUR nuclear research reactor core X-1 configuration by two ways; a deterministic way using WIMSD-CITVAP neutronic calculation in a fine calculation model, and a second way by a probabilistic method using MCNP5 model. The obtained results give the PPFs for each nuclear reactor core channel and the hot channel is then determined. Once this channel known, an asymptotic thermalhydraulic model is applied. This model gives the coolant and the cladding temperature evolutions in the hot channel for the calculated PPFs in the two presented models. All these results are compared with those obtained using the old PPF commonly used in such calculations. Results show that PPFs obtained by deterministic method are the most conservative. Thus, this last result is retained for a nuclear safety evaluation. An equation set-up is made for calculating the critical heat flux ratio. This ratio is verified to be greater than the limit of nuclear safety operating margin for NUR nuclear reactor. [ABSTRACT FROM AUTHOR]

Book review: Plokhy, Serhii. Chernobyl: The History of the Nuclear Catastrophe. New York: Basic Books, 2018. 423 pp. ISBN 9781541617087 Reviewed by Mariia Shuvalova

Hydrogen is increasingly being viewed as a significant fuel for future industrial processes as it offers pathways to zero emission. The UK sees hydrogen as one of a handful of low-carbon solutions for transition to net zero. Currently, most hydrogen production is from steam reforming of natural gas or coal gasification, both of which involve the release of carbon dioxide. Hydrogen production from mini decentralised grids via a thermochemical process, coupled with electricity production, could offer favourable economics for small modular reactors (SMRs), whereby demand or grid management as a solution would include redirecting the power for hydrogen production when electricity demand is low. It also offers a clean-energy alternative to the aforementioned means. SMRs could offer favourable economics due to their flexible power system as part of the dual-output function. This study objective is to investigate the critical performance parameters associated with the nuclear power plant (NPP), the cycle working fluids, and control-system design for switching between electricity and hydrogen demand to support delivery as part of a mini grid system for a reactor power delivering up to approximately 600 MWth power. The novelty of the work is in the holistic parametric analysis undertaken using a novel in-house tool, which analyses the NPP using different working fluids, with a control function bolt-on at the offtake for hydrogen production. The results indicate that the flow conditions at the offtake can be maintained. The choice of working fluids affects the pressure component. However, the recuperator and heat-exchanger effectiveness are considered as efficiency-limiting factors for hydrogen production and electricity generation. As such, the benefit of high-technology heat exchangers cannot be underestimated. This is also true when deciding on the thermochemical process to bolt onto the plant. The temperature of the gas at the end of the pipeline should also be considered to ensure that the minimum temperature-requirement status for hydrogen production is met.

We report an improved measurement of the neutrino mixing angle θ13 from the Daya Bay Reactor Neutrino Experiment. We exclude a zero value for sin22θ13 with a significance of 7.7 standard deviations. Electron antineutrinos from six reactors of 2.9 GW th were detected in six antineutrino detectors deployed in two near (flux-weighted baselines of 470 m and 576 m) and one far (1648 m) underground experimental halls. Using 139 days of data, 28909 (205308) electron antineutrino candidates were detected at the far hall (near halls). The ratio of the observed to the expected number of antineutrinos assuming no oscillations at the far hall is 0.944±0.007(stat.)±0.003(syst.). An analysis of the relative rates in six detectors finds sin22θ13=0. 089±0.010(stat.)±0.005(syst.) in a three-neutrino framework. © 2013 Chinese Physical Society and the Institute of High Energy Physics of the Chinese Academy of Sciences and the Institute of Modern Physics of the Chinese Academy of Sciences and IOP Publishing Ltd.

The selection of materials for the reactor pressure vessel manufacturing is a complex process that involves great responsibility because small differences in chemical composition can adversely affect the manufacturing process and the in-service behavior of the material. Thus, it is recommendable to perform previous materials pre-selection stages based on the state-of-the-art knowledge, integrating research results with standardized requirements and using simplifier materials selection methodologies like the stringency level method. To address this issue, an evaluation of the influence of chemical composition on the shift of the ductile-to-brittle transition temperature has been performed using the most used and consolidated prediction models that are R.G. 1.99 Rev.2, NUREG/CR-6551, and ASTM E 900-02. A proposal of maximum limits for copper, nickel, and phosphorous to avoid irradiation embrittlement has been presented to carry out the process. The results have been analyzed by using the stringency level methodology to support the decision process. To this end, a materials data collection has been carried out to analyze the requirements described by 20 different specifications of materials from first to fourth generation of light water reactors, covering the main designs of pressurized reactors from Western Europe, North America, Japan, and Russia. It can be concluded that more recently developed materials exhibit more stringent requirements than earlier developed materials. [ABSTRACT FROM AUTHOR]

The United Kingdom has declared a climate change policy of 100% reduction in carbon dioxide emissions by 2050. Efforts thus far have been limited solely to electricity generation methods. While progress has been admirable, effort now must be directed at the nation’s non-electrical energy use. Nuclear energy is an essential part of any energy future, since it is low-carbon, firm and supplies synchronous electricity; however the nation’s nuclear strategy to date has been erratic, costly and lacking in strategic oversight. A multitude of reactor designs are on offer for potential uptake, and decision-makers must have clarity of vision on what these systems must deliver before forming a strategy. Choosing between these systems, given the uncharted energy future faced by the UK is a daunting prospect. Generic feasibility assessment offers a tool for decision-makers to assist them in selecting the most suitable nuclear system for chosen future conditions. Generic feasibility assessment offers an alternative to traditional multi-attribute decision analyses, which can be confusing to even committed stakeholders when large numbers of attributes are weighted and compiled. Generic feasibility assessment forms part of a toolkit which will be of utility in achieving net zero by 2050, given the short time that remains.

Research activities of Vinca Institite have been based on two heavy water research reactors: 10 MW one, RA and zero power RB. Reactor RA was operational from 1962 to 1982. In 2010, spent fuel have been sent to the country of origin, and reactor now is in decommissioning. During operational phase of the reactor there were no recorded accidental releases into the environment just operational ones. Results of the environmental impact assessment, of the assumed emission of radionuclides, from the ventilation of nuclear reactor "RA" in Vinca, to the atmospheric boundary layer are presented in this paper. Evaluation was done by using the Gaussian straight-line diffusion model and taking into account characteristics of the reactor ventilation system, the assumed emission release of radioactivity (from the literature), site-specific meteorological data for six-year period and local topography around nuclear reactor, and corresponding dose factors for inventory of radionuclides. Based on the described approach, and assuming that the range of appropriate meteorological data for six year period for the application of described mathematical model is enough for this kind of analysis, it can be concluded that the nuclear reactor "RA", in the course of its work from 1962 to 1982, had no influence on the surrounding environment through the air above regulatory limits. [Projekat Ministarstva nauke Republike Srbije, br. III 45003]

Critical experiments using an insertion zone with FLiBe salt performed at the LR-0 reactor at Research Centre Řež (RC Řež) have been compared to application models for solid-fueled, fluoride-salt-cooled high-temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts using sensitivity and uncertainty (S/U) analysis techniques. These experiments support FHR and MSR advanced reactor concepts by informing on neutron spectral effects and nuclear data uncertainties related to fluoride salts. The FLiBe salt in the LR-0 experiments is from the Oak Ridge National Laboratory Molten Salt Reactor Experiment and is enriched to greater than 99.99% 7 Li. This work is part of a broader collaboration between the United States and the Czech Republic on MSR and FHR technology. Results from the S/U analyses for FHR and MSR models indicated the most significant potential source of eigenvalue bias due to nuclear data within the FLiBe salt is radiative capture in 7 Li. Other smaller but potentially significant contributions come from 19 F, 6 Li, and other 7 Li reactions. Similarity comparisons of the RC Řež LR-0 experiments and FHR and MSR application models indicate that the LR-0 FLiBe experiments could be useful as candidate benchmarks for low-enriched uranium–fueled application models. However, sensitivity differences observed included both spectral effects, due to the LR-0 reactor being moderated by water instead of graphite and fluoride salt, and sensitivity magnitude effects driven by the amount of FLiBe inserted and its location. New experiments with refined designs to increase the volume and importance of salt in the system could provide improved data. Results also demonstrate that salt systems using thorium and/or 233 U fuel will require additional experiments with relevant driver fuels. Increased contributions from the graphite moderator and 19 F and covariance between reactions such as 233 U fission and 233 U radiative capture indicate higher uncertainty contributions from the salt and significant differences in contributions from the fuel species due to underlying uncertainties in their nuclear data. [ABSTRACT FROM AUTHOR]

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The Loss Of Flow Accident (LOFA), with protection and without intervention of natural convection in the cooling process, is studied in the NUR nuclear research reactor for two accident configurations; a Fast Loss Of Flow Accident (FLOFA) and a Slow Loss Of Flow Accident (SLOFA). The accident is due to an inadvertent partial closure of coolant loop isolation valve. All the sequences of each accident are presented, namely the steady-state, the transient (SLOFA or FLOFA), the nuclear reactor emergency shutdown (SCRAM), the cooling of the decay heat and the contribution of 20% of flow to guarantee a good nuclear reactor core cooling. In order to be able to follow the thermalhydraulic accident down to the smallest detail, the equation set-up, based on the three conservation equations of continuity, momentum and energy, in three dimensions of space and one dimension of time, coupled to the turbulence model K − ω SST, is established. The models of the distribution of thermal power and decay heat are introduced into the mathematical model as well as the exponential flow loss equation. The boundary conditions for each sequence of the accident are fixed. The resolution of all the equations was carried out using Computational Fluid Dynamics (CFD) through FLUENT calculation code. It was possible to show the evolution of the temperature of the coolant and the cladding in the respective planes of each physical quantity for each time step of the accident, one second for the FLOFA and 2 s for the SLOFA. The evolution of the reactor power is presented before, during and after the SCRAM for each accident. After the SCRAM, the decay heat is taken into account and the average evolution of the temperature of the coolant and the cladding is presented as a function of time. The cooling of decay heat with no adding flow and with adding 20% flow was also studied as a function of time and the difference between the two approaches is presented. The FLOFA was studied for 2.2 s and the SLOFA for 25 s. The results obtained for the FLOFA and the SLOFA show that these accidents generates high temperatures but always remaining below the critical temperatures whether for the heat transfer fluid or for the cladding for the conditions and the data which have been retained for this work. [ABSTRACT FROM AUTHOR]

The current generation of nuclear reactors are evolutionary in design, mostly based on the technology originally designed to power submarines, and dominated by light water reactors. The aims of the Generation IV consortium are driven by sustainability, safety and reliability, economics, and proliferation resistance. The aims are extended here to encompass the ultimate and universal vision for strategic development of energy production, the "perpetuum mobile"--at least as close as possible. We propose to rethink nuclear reactor design with the mission to develop an innovative system which uses no fresh resources and produces no fresh waste during operation as well as generates power safe and reliably in economic way. The results of the innovative simulations presented here demonstrate that, from a theoretical perspective, it is feasible to fulfil the mission through the direct reuse of spent nuclear fuel from currently operating reactors as the fuel for a proposed new reactor. The produced waste is less burdensome than current spent nuclear fuel which is used as feed to the system. However, safety, reliability and operational economics will need to be demonstrated to create the basis for the long term success of nuclear reactors as a major carbon free, sustainable, and applied highly reliable energy source. [ABSTRACT FROM AUTHOR]

Neutronic calculations are performed for the X-1 configuration of NUR nuclear research reactor core, followed by a CFD thermalhydraulic modeling of the reactor core hot channel. The neutronic study is established by considering the nuclear reactor core as being composed of homogenized zones. Calculated power densities in the fuel elements are used to identify the hottest channel in the nuclear reactor core and to determine the power peaking factors. To do this, the WIMSD-CITVAP neutron calculation line was used. This was followed by a 3D thermalhydraulic modeling. The mathematical model is based on equations of conservation of mass, momentum and energy. These equations are coupled to a k − ω SST turbulence model. The calculated power peaking factors are introduced into the power distribution equation which is applied to the outer area of the fuel meat. This equation set-up is solved, on a numerical domain mesh, using the FLUENT CFD code. In order to validate the method developed in this work, an application was made and the temperature distribution profile of the cladding was plotted. A comparison with a published work has shown that the calculated relative differences do not exceed 2.4 % and that the temperature profile obtained in this work is the most conservative. A parametric study was conducted to confirm the physically valid behavior of the model developed in this work and the obtained results also confirm that Onset of Nucleate Boiling (ONB) and cladding melting cannot occur under the conditions of the present study. [ABSTRACT FROM AUTHOR]

Abstract: A conceptual design is presented for a hybrid sulfur process for the production of hydrogen using a high-temperature nuclear heat source to split water. The process combines proton exchange membrane-based SO2-depolarized electrolyzer technology being developed at Savannah River National Laboratory with silicon carbide bayonet decomposition reactor technology being developed at Sandia National Laboratories. Both are part of the US DOE Nuclear Hydrogen Initiative. The flowsheet otherwise uses only proven chemical process components. Electrolyzer product is concentrated from 50wt% sulfuric acid to 75wt% via recuperative vacuum distillation. Pinch analysis is used to predict the high-temperature heat requirement for sulfuric acid decomposition. An Aspen Plus™ model of the flowsheet indicates 340.3kJ high-temperature heat, 75.5kJ low-temperature heat, 1.31kJ low-pressure steam, and 120.9kJ electric power are consumed per mole of H2 product, giving an LHV efficiency of 35.3% (41.7% HHV efficiency) if electric power is available at a conversion efficiency of 45%. [Copyright &y& Elsevier]

• Developed a detailed model of the ML-1 reactor for use with MCNP. • Compared ML-1 MCNP model to experimental data obtained between 1962 and 1965. • Improved design with advanced fuel forms to achieve 3 years between refueling cycles. • Developed a cost-estimate for this micro-reactor concept using Congressional records. This work examines a mobile, land-based micro-reactor concept based on the mobile low power reactor prototype 1 (ML-1). A detailed model of the ML-1 micro-reactor was created in MCNP® and compared with experimental data from multiple demonstrations conducted between 1962 and 1965. The computed values agree with experimental data, evidenced by an obtained k eff of 1.10705 ± 0.00104 versus an expected value of 1.102 ± 0.007. A low cost-estimate of $71.1 million and high cost- estimate of $291.3 million in today's dollars for a micro-reactor design based on the ML-1 concept was obtained using program costs captured in Congressional legislative records and the average inflation rate. Improvements were made to the ML-1 design using advanced fuel materials to meet contemporary design requirements. It has been demonstrated that a modernized reactor configuration utilizing fuel in uranium carbide form, with 19.75% U-235 enrichment, and helium coolant can support steady state operation at 3 MWth for 3 years. [ABSTRACT FROM AUTHOR]

A complete computational simulation based on the Angra dos Reis nuclear reactor neutrino experiment was made. The simulation includes the implementation of a cylindric neutrino detector with walls used like veto against external background, all simulated in Geant 4 and GLG4sim. A feasibility study about the measurement of the Weinberg angle in this experiment was made. [ABSTRACT FROM AUTHOR]

Despite renovation of the D2O facility, tritium concentrations in the condensates of reactor room air showed tens of Bq mL-1 before venting resumption on July 1997. This suggested the presence of tritium sources in the research reactor-containment building. Au investigation was therefore initiated to locate the source and determine the distribution of tritium in the containment building. Air monitoring in the working area using a dish of water placed in the building suggested that the source of tritium was near the reactor core. Monitoring exhaust air from the two facilities (a cold neutron source and a D2O tank) showed high specific activity on the order of 105 Bq mL-1, suggesting the presence of tritium in condensates near the reactor core. The major concern was whether the leakage of liquid deuterium (4 L) and heavy water (2 × 10³ L) used as a moderator had occurred. The concentration of tritium in condensates has not increased over the past few years in either the exhaust line or working area, and the deuterium itself has not been found in the surrounding environment. The concentration of tritium measured using an ionization chamber after 41Ar decay was dependent on the thermal output of the research reactor, indicating that the tritium was produced by the irradiation process within shielding/moderator materials or cover gas with neutrons. [ABSTRACT FROM AUTHOR]

In this article an overview of the historical development of light water reactor health physics programs is presented. Operational health physics programs have developed and matured as experience in operating and maintaining light water reactors has been gained. Initial programs grew quickly in both size and complexity with the number and size of nuclear units under construction and in operation. Operational health physics programs evolved to face various challenges confronted by the nuclear industry, increasing the effectiveness of radiological safety measures. Industry improvements in radiological safety performance have resulted in significant decreases in annual collective exposures from a high value of 790 person-rem in 1980 to 117 person-rem per reactor in 2002. Though significant gains have been made, the continued viability of the nuclear power industry is confronted with an aging workforce, as well as the challenges posed by deregulation and the need to maintain operational excellence. [ABSTRACT FROM AUTHOR]

The upper limits for the rate of release of radionuclides into the atmosphere, i.e., the "derived release limits," are calculated for the Greek Research Reactor (GRR-1) in order to determine possible operational schemes compatible with the effective dose limits for the general population. GRR-1 is located at the northwestern foot of Hymettos Mountain and at the eastern border of the urbanized area of Athens basin. Due to the topographic complexity of the region, the meteorological and atmospheric dispersion calculations were based on a numerical modeling system that is especially designed to work over irregular terrains by using a prismatic unstructured grid. The calculation of derived release limits was made using guidelines and methods that conform to the system of dose limits prescribed by the European radiation protection regulations. [ABSTRACT FROM AUTHOR]

In situ gamma spectroscopy was used to measure 41Ar released into the laboratory due to the activation of air in the irradiation tubes of SLOWPOKE reactor during routine neutron activation analysis. The data obtained were used to predict the 41Ar distribution in the laboratory for various operating conditions of the reactor. The dose received by the analyst from the immersion in 41Ar was calculated to be 2.36 µSv γ-1, which is approximately 1% of the normal background exposure. [ABSTRACT FROM AUTHOR]

The field history and current status of internal dosimetry is reviewed in this article. Elements of the field that are reviewed include standards and models, derivation of dose coefficients and intake retention fractions, bioassay measurements, and intake and dose calculations. In addition, guidance is developed and provided as to the necessity of internal dosimetry for a particular facility or operation and methodology for implementing a program. A discussion of the purposes of internal dosimetry is included as well as recommendations for future development and direction. [ABSTRACT FROM AUTHOR]