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Precise normalization of ToF yield measurements of neutron induced fission in fissile targets is challenging, but the appropriate normalization is also of critical importance for nuclear energy and criticality safety, among other applications. A typical normalization relies on the Thermal Neutron Constants (TNC) recommended at the thermal point (fission σ f 0 and capture σ γ 0 thermal cross sections) by the Neutron Standards. However, many ToF experiments do not collect data down to the thermal energy and use normalization cross-section integrals defined at different arbitrary energy intervals. Normalization fission cross-section integrals I 3 are recommended in between-valleys energy regions 8.1–14.7 eV, 7.8–11 eV, 9–20 eV, and 11.7–19.5 eV for fissile targets 233U, 235U, 239Pu, and 241Pu with values equal to 689.0(10.8), 245.7(4.1), 1059(6), and 1378(33) b ⋅ e V , respectively. The 235 U normalization integral I 3 derived in this work of 245.7(4.1) b ⋅ e V is in good agreement within quoted uncertainties with the Neutron Standards value of 247.5(3.0) b ⋅ e V. Additional cross-section integrals I 1 in the thermal region from 20–60 meV are derived to fix both the normalization and the slope of the fission cross section at the thermal point providing additional constraints for R-matrix evaluations of experimental fission yields. Ratios σ f 0 / I 1 and I 3 / I 1 feature very low uncertainty due to the strong positive correlations between the numerator and the denominator and are comprehensively derived for the first time. Integral ratios I 3 / I 1 of 39.31(54), 13.08(20), 41.65(22), and 40.46(85) are recommended as reference for fissile targets 233U, 235U, 239Pu, and 241Pu, respectively. Similarly, ratios σ f 0 / I 1 of 30.40(18), 31.22(12), 29.58(7), and 29.95(35) are also recommended as reference for corresponding fissile targets. These recommended ratios can be used by experimentalists to normalize their measured thin-target fission yield data, but also by evaluators to constrain their R-matrix analysis that use multiple sets of measured data renormalized as recommended in this work. An evaluation of thermal (n,f) cross sections σ f 0 presented in this work for all fissile targets is in excellent agreement with the Thermal Neutron Constants defined in the Neutron Standards within one-sigma uncertainty. Such agreement guarantees the reliability of our new evaluation of reference integrals based on the whole database of consistent experimental TOF data. [ABSTRACT FROM AUTHOR]

The 9/11 attacks prompted the development of an international nuclear security regime. States are expected to adopt legislation and institutionalize measures to ensure cooperation among stakeholders and create their own national nuclear security regimes. This article evaluates the steps taken by Turkey, a newcomer in nuclear energy. It argues that while Ankara has acted in line with its traditional conception of security, this is not enough. The inadequate cooperation and coordination among stakeholders, and the insufficient development of a security culture specifically oriented to nuclear energy, leave the country vulnerable. The analysis explains the vulnerabilities and incomplete tasks, and proposes actions necessary for Turkey to create a sound national nuclear security regime. [ABSTRACT FROM AUTHOR]

In this article, we present the results of the creation of an information-analytical system (IAS) developed as a tool for research and prediction of the behavior of the corium simulating a core melt of a nuclear reactor in the experiments. This was carried out to justify the safety of the use of nuclear energy according to specific requirements of a customer. The data loaded into the IAS were obtained on the basis of the National Nuclear Center of the Republic of Kazakhstan (NNC) as a result of modeling the processes occurring during severe accidents at nuclear installations. The experimental data obtained at the benches of NNC and characterizing the processes occurring during the interaction of the core melt of the nuclear reactor with coolants, structural materials and structural elements of nuclear reactors were classified and structured for the convenience of analyzing and processing the data obtained. The structure of the created IAS is shown and the analytical block is described in detail. [ABSTRACT FROM AUTHOR]

The recovery and extraction of uranium is vital to ensure sustainability of nuclear energy production and safety of environment. Herein, we report a facile method to synthesise diglycolamic-acid-functionalized iron oxide@titanium dioxide magnetic nanocomposite (Fe 3 O 4 @TiO 2 -DGA) as an efficient adsorbent for effective removal of uranium (VI) from aqueous medium. Fe 3 O 4 @TiO 2 -DGA showed high uranium adsorption performance, with a sorption capacity of 371.8 mg g–1 at pH 6.0 and reached equilibrium within 40 min. The adsorption isotherm fits well with Langmuir and the kinetics follows pseudo-second-order model, respectively. The plausible mechanism of U(VI) adsorption involves the electrostatic interaction between the diglycolamic acid and hydrolysed uranium species. Thermodynamic studies revealed that the adsorption was spontaneous and exothermic in nature. The U(VI)-loaded Fe 3 O 4 @TiO 2 -DGA could be conveniently separated from aqueous solutions using an external magnet, and further U(VI) is desorbed by adding dil. HCl. Fe 3 O 4 @TiO 2 -DGA showed exceptional selectivity for U(VI) in presence of various coexisting metal ions and it can be effectively reused for more than 5 times with negligible change in sorption capacity. Thus, Fe 3 O 4 @TiO 2 -DGA can be considered as an effective and potential sorbent for U(VI) remediation from aqueous solutions. [Display omitted] • Fe 3 O 4 @TiO 2 -DGA is synthesized through facile method for enhanced adsorptive removal of U(VI) from aqueous solutions. • Fe 3 O 4 @TiO 2 -DGA can selectively extract U(VI) with a maximum adsorption capacity (q max) of 371.8 mg g−1 at pH 6.0. • Equilibrium adsorption is achieved with minimum contact time of 40 min • The nanocomposite exhibited significant reusability even after several adsorption cycles. [ABSTRACT FROM AUTHOR]

Ensuring the safety of nuclear energy necessitates proactive measures to prevent the escalation of severe operational conditions. This article presents an efficient and interpretable framework for the swift identification of abnormal events in nuclear power plants (NPPs), equipping operators with timely insights for effective decision-making. A novel neural network architecture, combining Long Short-Term Memory (LSTM) and attention mechanisms, is proposed to address the challenge of signal coupling. The derivative dynamic time warping (DDTW) method enhances interpretability by comparing time series operating parameters during abnormal and normal states. Experimental validation demonstrates high real-time accuracy, underscoring the broader applicability of the approach across NPPs. [ABSTRACT FROM AUTHOR]

Safety culture is a very important tool to improve safety at nuclear energy facilities. It develops abilities of personnel to efficiently mitigate potential threats to nuclear and radioactive material, promoting correct attitude and behavior, as a consequence of which the personnel gain more stringent and cautious approach to their duties in the field of physical safety. In the companies with high culture of physical nuclear safety, the chance of a malicious act by personnel is hardly probable owing to persuasion that nuclear safety is important as well as owing to the factor of vigilance and professional compliance with the practices of physical nuclear safety. This article discusses the issue of development of safety culture in the Republic of Belarus in 2020–2021. This time period coincided, on one hand, with the commissioning of the first nuclear power unit in Ostrovets and, on the other hand, with the spread of COVID-19. The role of safety culture has increased as mentioned in some regulations approved in 2020. The role, the functions, and the requirements for safety culture have been formulated not only for the Belarussian NPP but also for the companies operating sources of ionizing irradiation. The current times are dealing with the spread of COVID-19. Introduction of sanitary norms added corrections to arrangement of measures required for formation and improvement of safety culture. This article presents the approach and summarizes the measures developed by Joint Institute for Power and Nuclear Research—Sosny, National Academy of Sciences of Belarus, aiming at a higher level of safety culture. [ABSTRACT FROM AUTHOR]

A dual-coolant integrated experimental facility named DRAGON-V has been developed at the Institute of Nuclear Energy Safety Technology, Hefei Institute of Physical Science, Chinese Academy of Sciences, for the key technology research and performance evaluation of candidate liquid lithium-lead (PbLi) blanket of hydrogen fusion reactors. The loop is composed of a material test sub-loop and thermal-hydraulic test sub-loop, the design parameters are PbLi inventory 20 tons, PbLi temperature up to 550 °C, the maximum PbLi flow rate up to 40 kg/s. A novel cold trap system is designed to remove the suspended and crystalized impurities in PbLi fluid with three cooling zones and cross row arrangement of rod bundle filter elements. The paper describes the loop itself and its major components, initial loop testing, flow and measurement diagnostics and current experiments. The obtained test results of the loop and its components have demonstrated that the new facility is fully functioning and ready for experimental studies of material corrosion with/without a magnetic field, magnetohydrodynamic (MHD) effect, purification, heat and mass transfer phenomena in PbLi flows and can also be used in mock-up testing in conditions relevant to fusion applications. • A lithium-lead facility with 40 kg/s mass flow and 2–5 T magnetic field was built. • The composition and performance of the device are described and tested in detail. • A novel cold trap system with three cooling zones has been designed and proposed. • The safety analysis of loss of flow accident and loss of heat sink was investigated. [ABSTRACT FROM AUTHOR]

• User-defined function (UDF) based on the Gaussian moving heat source is applied. • A 3D thermal-stress coupling numerical model is established. • The variation law of the thermal stress in the spent fuel storage tank is obtained. • The main factor affecting stress is the variation in temperature gradient. Spent fuel storage is a necessary part of the entire lifecycle to achieve intrinsic safety in the application of nuclear energy as a clean energy source, and should be adequately investigated. Based on a Gaussian stochastic heat source, this paper presents a three-dimensional coupled numerical model of the heat-stress multiphysics field for the dry storage cask. Comparison with the literature model is carried out to verify that the error is less than 10%, and the temperature field and stress field distribution law of the storage cask under different initial temperatures of the spent fuel rods are studied and discussed. The raise of the initial temperature enhances the heat transfer and the change of the temperature gradient across the shell is the main cause of the stress change. This study provides a reference for the subsequent design study of spent fuel storage casks. [ABSTRACT FROM AUTHOR]

Thallium doped Cesium Iodide (CsI:Tl) radioluminescent microcrystals are promising candidate for scintillation applications that include medical imaging, nuclear safety and high energy particle detection applications. This paper introduces, for the very first time, an antisolvent recrystallization method to synthesize microcrystals of CsI:Tl. The antisolvent recrystallization method is useful for synthesizing microcrystals at room temperature under atmospheric pressure and humidity conditions with high product yield approaching ∼90 %. The polycrystalline nature of the CsI:Tl microcrystals is confirmed from the X-ray diffraction analysis, wherein the crystallite size is found to be in the range of 20–25 nm. The scanning electron microscopy revealed the size of microcrystals in the range of 1–13 μm. The doping concentration of thallium (with 0.06, 0.10, 0.20, 0.50 and 1.00 mol%) strongly affects the luminescence yield. At higher concentration of doping, a shift towards long wavelength is observed. The strongest luminescence yield is observed in the sample of 0.2 mol% Thallium doping, which is around 76 % of luminescence compared to single crystal of CsI:Tl. This work provides a low-cost alternative to the growth of radioluminescent microcrystals in comparison with the existing single crystal growth approaches. CsI (0.2 mol% Tl) microcrystals (A) Normal light (B) Under the exposure of UV light. [Display omitted] • Novel method for CsI:Tl microcrystals synthesis in ambient conditions. • CsI:Tl (0.2 mol% Tl) yields maximum luminescence efficiency. • Achieves 76 % luminescence of CsI:Tl single crystals. • Cost-effective alternative for radioluminescent microcrystal growth compared to single crystals. [ABSTRACT FROM AUTHOR]

After the Fukushima disaster of March 2011, the government of Japan created new regulations for nuclear power and nuclear reprocessing (i.e. the recycling of spent fuel). The new compliance review process deserves credit for closing some safety gaps, and if the owners of the Rokkasho reprocessing plant, a major facility under construction for more than two decades now, implement adequate measures and operate the facility as expected based on the new requirements, the plant's safety could certainly improve. However, potential risks have thus far gone unaddressed by the new review process, and unless the government deals with these problems, it will not be able to ensure maximum safety. [ABSTRACT FROM AUTHOR]

The lessons learned from the Fukushima Daiichi accident have focused on preventive measures designed to protect nuclear reactors, and crisis management plans. Although there is still no end in sight to the accident that occurred on March 11, 2011, how engineers have handled the aftermath offers new insight into the capacity of organizations to adapt in situations that far exceed the scope of safety standards based on probabilistic risk assessment and on the comprehensive identification of disaster scenarios. Ongoing crises in which conventional resources are lacking, but societal expectations are high, call for “engineering thinking in emergency situations.” This is a new concept that emphasizes adaptability and resilience within organizations—such as the ability to create temporary new organizational structures; to quickly switch from a normal state to an innovative mode; and to integrate a social dimension into engineering activities. In the future, nuclear safety oversight authorities should assess the ability of plant operators to create and implement effective engineering strategies on the fly, and should require that operators demonstrate the capability for resilience in the aftermath of an accident. [ABSTRACT FROM PUBLISHER]

Radioactive fire following nuclear accidents causes resuspension of radioactive contamination that can spread over large areas. Although many papers focus on particle deposition and resuspension, a limited number of studies model its suppression. Even less data is available for the comparison of emissions across particle size, surface, and mode of resuspension. In this study, cerium dioxide surrogate for internal hazards such as radionuclide particle matter was aerosolized into a large chamber and subsequently resuspended after sedimentation. Resuspension was induced by common modes of human movement including walking, wind, and driving on common surfaces including polyvinyl chloride (PVC), artificial grass (turf), and concrete in a 5.8 m long, 3.7 m wide and 3 m high tent. The resuspension was experimentally measured for two particle sizes (1 and 10 μm) at three different heights (0.3 m, 1.2 m, and 2.4 m) and compared across size, surface, and mode of resuspension. Particle size distribution was measured using an aerodynamic particle sizer and the resuspended particles were collected with the wetted wall cyclone aerosol collector and quantitated using inductively coupled plasma mass spectrometry. The testing was repeated with the F-500 fire retardant application to study its effect on resuspension. In this study, artificial grass flooring was found to generate the most resuspension, with walking on turf for the 1 μm particle testing yielding the highest factor overall. Concrete had the lowest resuspension, showing variations in the size distribution of aerosol as a function of height from the source resuspension factors. The application of fire retardant was found to significantly reduce surrogate resuspension, regardless of resuspension method or testing surface. The results provide impactful information regarding the different modes of resuspension including human transport (walking and driving) and wind for the resuspension of radionuclide surrogate particles deposited onto common surfaces and the effect of fire retardant application on their resuspension, leading to higher levels of safety for nuclear energy plants and facilities. • Effect of fire retardant on surrogate resuspension following nuclear accidents. • CeO 2 particles in two sizes resuspended by three modes after sedimentation on three surfaces. • Resuspended particles collected with the WWC and quantitated using ICP-MS. • Highest resuspension factors found for artificial grass, lowest for concrete surfaces. [ABSTRACT FROM AUTHOR]

Due to its radiotoxicity tritium, especially in its oxide form, represents the characteristic hazard of a Tritium Removal Facility (TRF). Consequently, the risk of uncontrolled releases of tritiated water (HTO) into the free atmosphere is considered to be the characteristic risk (Rc) of these non-reactor nuclear installations. The Rc max of a TRF will be characterized by the maximum and credible probability and by the maximum possible consequences of the most unfavorable accident scenario, usually meaning uncontrolled release of tritium as oxide in the outer atmosphere, by failing of all defense-in-depth barriers. Locating it in the acceptable zone of the risk matrix would be the best confirmation of the project quality from the perspective of life and environment security, generating a consolidation of public trust in the safety of nuclear energy. The paper presents a semi-quantitative determination of the maximum characteristic risk for the TRF pilot plant which can be applied or can serve as an approach for other tritium extraction facilities connected to fission CANDU (Canada Deuterium Uranium) or fusion type reactors, like ITER (International Thermonuclear Experimental Reactor) or DEMO (DEMOnstration Power Plant). The advantage of this method is that it requires a low consumption of man-hours and can be carried out from the design phase with minimum input data. The results allow the verification of compliance with national legal provisions, of the location choice and the development of the emergency plan in case of a maximum credible accident. [ABSTRACT FROM AUTHOR]

In the two years since the Fukushima Daiichi nuclear accidents, countries around the world have responded to the Japanese catastrophe in a variety of ways. Broadly speaking, these policy responses fall into three categories: Within one group, countries have decided to turn away from nuclear power, either through a gradual phase-out or by abandoning plans for programs that had not yet begun. In the second group of countries, government leaders who were otherwise inclined to continue pursuing nuclear power were forced by protests and public opinion to change their policies. And in the third group, countries have reiterated their commitment to nuclear power. The author describes examples of countries within each of these categories. He identifies several common elements that seem to be at play in countries that are staying the course on nuclear power, including: hasty dismissals by government officials of the Fukushima accident’s applicability to their own countries, pronouncements about the safety of deployed or proposed reactor designs, propaganda campaigns, international financing for reactor construction, and an emphasis on projected growth in energy demand. [ABSTRACT FROM PUBLISHER]

Could nuclear power be rapidly expanded on a global scale? There are a number of practical limiting factors, including site availability and acceptability, nuclear waste disposal issues, and the risks of accidents and proliferation. But there are also a variety of resource limitations. One particular resource limitation that has not been clearly articulated in the nuclear debate thus far is the availability of the relatively scarce metals used in the construction of the reactor vessel and core. While this scarcity is not of immediate concern, it would present a hard limit to the ultimate expansion of nuclear power. This limit appears to be a harder one than the supply of uranium fuel. An increased demand for rare metals—such as hafnium, beryllium, zirconium, and niobium, for example—would also increase their price volatility and limit their rate of uptake in nuclear power stations. Metals used in the nuclear vessel eventually become radioactive and, on decommissioning, those with long half-lives cannot be recycled on timescales useful to human civilization. Thus, a large-scale expansion of nuclear power would reduce “elemental diversity” by depleting the world’s supply of some elements and making them unavailable to future generations. [ABSTRACT FROM PUBLISHER]

To enhance the safety of nuclear power, the focus of researchers all around the world has recently mainly objected on the development of Accident Tolerant Fuels. Especially the Chromium coating of current Zirconium based cladding has been widely suggested and discussed for its immense positive effect on overall cladding properties. Nevertheless, it was observed that during the first stage of the Loss of Coolant Accident, cracks appear in the Cr coating due to its inability to tolerate higher plastic strain. Therefore, experimental methodology used in this article focuses on testing fuel cladding with damaged Cr coating after the high-temperature transient. The impact of cracks on degradation of cladding mechanical properties was observed using optical microscopy, ring compression test, microhardness, and evaluating hydrogen content and weight gain. [ABSTRACT FROM AUTHOR]

The article explores the introduction of Probabilistic Risk Assessment (PRA) for nuclear energy in the two German states, the FRG and the GDR since the late 1960s. We argue that PRA - which promised to make potential dangers associated with the new technology calculable, comparable and seemingly controllable by reducing them to numerical terms - is best understood as an evidence practice, aiming to (re-)establish intersubjective agreement on nuclear safety through quantification. As such, the introduction of PRA was from the beginning also a political question, tied to the destabilization of alternative evidence practices. While in both the FRG and the GDR, the relativization of the promise of absolute safety inherent in the new method proved problematic, this was an even bigger obstacle in the socialist East. Although PRA ultimately failed to (re-)establish a societal consensus on nuclear energy in Germany, its institutionalization shaped the societal discourse on dangerous technologies. [ABSTRACT FROM AUTHOR]

In this interview, the physicist Brice Smith comments on the risks of using nuclear power as an alternative energy source. Smith discusses the risks involved in nuclear power and the problems of developing public awareness of those risks. Smith also discusses the economics of nuclear power and the challenges to financing a nuclear power plant.

Design evaluation of nuclear components using numerical methods has typically been focused on nominal geometries and conditions. In reality, the geometry of reactor components and operating conditions can differ from the design on the drawing table. Increasing understanding and safety of nuclear energy systems requires investigating these more realistic conditions as well. An example is the bow of fuel assemblies, a deformation that occurs after some residence time as a consequence of thermal and irradiation effects. A paradigm shift from purely deterministic simulations toward simulations involving stochastic processes can help to meet this demand. One of the aspects in the design assessment of nuclear reactors is flow-induced vibration (FIV). A lot of experimental effort has already been allocated to FIV phenomena in the past, and more recently also numerical approaches have been applied. However, it was not clear to what extent vibration characteristics would change for a geometry different from the design drawing. In this study the effect of a random bow deformation on the vibration behavior of a flexible tube in a rigid bundle is investigated using fluid–structure interaction simulations. As the (stressless) geometry presumably has a stochastic nature, the modal characteristics should be treated accordingly. For this uncertainty quantification of the modal characteristics, the generalized Polynomial Chaos method was applied. First, only the steady flow through the bundle was considered, without simulating the vibration. This sufficiently reduced the computational cost to apply the Monte Carlo method, which was used to validate the more efficient Polynomial Chaos method. This latter method was found adequate to predict the stochastic modal characteristics of the deformed, flexible tube. • Bow deformation of fuel pin modeled as two independent stochastic input variables. • Uncertainty quantification with generalized Polynomial Chaos (gPC) and Monte Carlo. • Effect of random bow on flow propagated using gPC and validated using Monte Carlo. • FSI simulations for gPC samples to evaluate first mode frequency and damping ratio. • Uncertain frequency and damping due to random bow determined using gPC method. [ABSTRACT FROM AUTHOR]

Highlights • Chinese nuclear safety regulatory framework and ITER licensing lessons were reviewed. • Public dose limit of 0.1 mSv/a and same safety limit of ITER was proposed for CFETR. • Tritium emission limit of 5–33 g/a was proposed for CFETR. • Public radiation dose and nuclear emergency were estimated and discussed for CFETR. Abstract Since Chinese fusion engineering test reactor (CFETR) stepped into the engineering design phase, it was significant to discuss the environmental radiation issues considering both Chinese radiation safety regulations and lessons learned from international thermonuclear experimental reactor (ITER) being constructed in France. In this study, firstly, the legal and regulatory framework of nuclear safety in China and ITER licensing process lessons were reviewed. Secondly, public dose limit was proposed for CFETR. Thirdly, safety limit and radioactive source terms were discussed and proposed for CFETR. Fourthly, public radiation dose was estimated, tritium emission limit was proposed under normal operation from the perspectives of tritium emission limit for nuclear power plants (NPPs) in China and reverse value based on public dose limit. Finally, potential consequences of environmental radiation were discussed under accidental conditions. Nuclear emergency countermeasures including no evacuation criteria needed or not were proposed. The preliminary environmental radiation considerations were proposed for CFETR: 1) It had better invite the experts from Chinese national nuclear safety administration (NNSA) to participate in the engineering design, which can make the related persons understand the fusion safety characteristics as early as possible, so as to further speed the licensing process; 2) 0.1 mSv/a was proposed as the public dose limit of CFETR under normal operation; 3) The same ITER safety limit was proposed to be set for CFETR from the consideration of both comparable size of vacuum vessel (VV) and hydrogen/dust explosion; 4) A tritium emission limit of 5–33 g/a was proposed for CFETR; 5) The "no off-site emergency response" was suggested as one of the safety goals during the design of CFETR, however, "off-site emergency measures" were still proposed to be deployed during the operation of CFETR. It was expected the environmental radiation considerations could provide a reference to speed the construction of CFETR on the premise of environment friendly and public safety. [ABSTRACT FROM AUTHOR]

Pebble bed reactors (PBRs) can improve the safety and economics of the nuclear energy production. PBRs rely on TRIstructural-ISOtropic (TRISO) fuel pebbles for enhanced fission product retention. Accurate characterization of individual fuel pebbles would enable the validation of computational models, efficient use of TRISO fuel, and improve fuel accountability. We have developed and tested a new neutron multiplicity counter (NMC) based on 192 boron coated straw (BCS) detectors optimized for 235 U assay in TRISO fuel. The new design yielded a singles and doubles neutron detection efficiency of 4.71% and 0.174%, respectively, and a die-away time of 16.7 μ s. The NMC has a low intrinsic gamma-ray detection efficiency of 8. 71 × 1 0 − 8 at an exposure rate of 80.3 mR/h. In simulation, a high-efficiency version of the NMC encompassing 396 straws was able to estimate the 235 U in a pebble with a relative uncertainty and error both below 2% in 100 s. • TRISO fueled-pebble characterization. • Boron-coated straw multiplicity counter. • Active interrogation of TRISO fuel. [ABSTRACT FROM AUTHOR]

Destruction of critical nuclear infrastructure would have a debilitating effect on national public health, safety, national economy and security. For this reason, analysts perform safety risk analyses on the performance of the nuclear system to quantify and understand the nature of unwanted events. Since the world has gone through many changes after the terrorist attacks of 9/11, nuclear security risk analysis also became a necessity. To date, the safety and security risk analyses have been done separately without a combined evaluation. Study results are presented for three types of risk analyses for a pure security initiating event, pure safety initiating event, and a combined analysis of safety-security risk for either a security or safety initiating event. The pure security risk analysis uses adversary sequence diagram and pathway analysis to calculate the initiating security event frequency of a successful adversary attack. The pure safety analysis represented a series of natural (random) safety system component failure events for which a safety system failure frequency was calculated using SAPHIRE probabilistic risk analysis code. On the other hand, the combined safety-security analysis considered a security initiating event followed by safety system failure or vice versa. The main outcome of the comparative study of three different types of risk analyses is that pure safety risk evaluation without considering the possibility of a simultaneous security attack would underestimate the risk value. Failure frequency due to a security event should be combined with the safety system failure analysis for a meaningful risk analysis and the Estimate of Adversary Sequence Interruption (EASI) model can be employed for this purpose. The usefulness of a combined safety-security risk analysis is demonstrated through a case study for the spent fuel storage pool facility. [ABSTRACT FROM AUTHOR]

We studied the capability of an annular, prismatic HTGR to remove decay heat passively. The purpose of the study was to obtain the design parameters relationship of the annular, prismatic HTGR with passive decay heat removal depending on power density profile and to compare them with those for solid cylinder one. The results showed that the safety feature of the annular reactor is improved a lot compared with that of cylinder one. The safety margin could be increased further by flattening the power density profile. Then fundamental neutronic analysis was performed for the annular reactor whose design parameters are obtained from the condition. [ABSTRACT FROM AUTHOR]

We analyze the availability of the Safety Programmable Logic Controller (SPLC) having multiple redundant architectures. In the SPLC, input/output and processor module are configured as triple modular redundancy (TMR), and backplane bus, power and communication modules are configured as dual modular redundancy (DMR). The voting logics for redundant architectures are based on the forwarding error detection. It means that the receivers perform the voting logics based on the status information of transmitters. To analyze the availability of SPLC, we construct the Markov model and simplify the model adopting the system unavailability rate. The results show that the fault coverage factor should be ≥0.8 and Mean Time To Repair (MTTR) should be ≤100 h in order to satisfy the requirement that the availability of the safety grade PLC should be ≥0.995. Also we evaluate the availability of SPLC comparing to other PLCs such as simplex, processor DMR (pDMR) and independent TMR (iTMR) PLCs used in the existing nuclear safety systems. The availability of SPLC is higher than those of the simplex, pDMR but is lower than that of iTMR for one month which is the periodic off-line test and inspection. That’s why the number of redundant modules used in PLC is more dominant to increasing the availability than the number of fault masking methods such as voting logics used in PLC on the assumption that operation time is in the early stage. But the availability of iTMR, which has many redundant modules but has only a voting logic fast decrease and eventually is the lowest after 8000 h. Also if the MTTR of each module in PLC is required to be increased to 200 h, the availability of SPLC would be better than iTMR. [ABSTRACT FROM AUTHOR]

Safety integrity level (SIL) verification is a critical step in safety lifecycle of safety-related systems (SRS). Introducing redundancy into SRS raises two issues: voting group configuration and common cause failures (CCF). In order to minimize CCF, diverse redundancy is widely adopted by SRS. However, in the past, almost all attention of SIL verification has been paid to identical redundancy, this is reflected in IEC 61508, ISA-TR84.00.02 and scientific literatures. Therefore, a novel method for SIL verification of SRS with diverse redundancy based on system degradation is proposed. Key idea of the method is to calculate average probability of dangerous failure on demand ( PFD G ) at each stage of system degradation, which is caused by failures of redundant channels. To validate proposed method, it has been applied on safety shutdown system of Nuclear Power Control Test Facility, and numerical result is compared with FTA and FRANTIC model. Sensitivity studies and comparison of numerical results indicate that the method has very good consistency with FTA and FRANTIC model. Moreover, two sets of general formulae for PFD G of any M oo N (D) group with diverse redundancy are provided. From engineering practice point of view, it makes SIL verification process simpler. [ABSTRACT FROM AUTHOR]

We explore the abilities of the Singular Spectrum Analysis (SSA) to characterize and denoise discrete acoustic emission signals. The method is first tested on simulated data for which different types and levels of noise are considered. It is then applied on real data recorded from nuclear safety experiments. The results show an excellent ability of the SSA to characterize the corrupted signal and to detect structural changes, even for low signal-to-noise ratio. For denoising purposes, the quality of the results depends mainly on the separability between the source signal to be estimated and the noise. However, whatever the case, the main components of the source signal are clearly identified when the components associated with the noise are removed. [ABSTRACT FROM AUTHOR]

In today’s world there is more than one point of view and different debates about how to determine and what constitutes the concept of nuclear security. A narrow attitude, which we can not agree with, is that it is an area that deals with issues of the model of physical security of nuclear materials. A broader interpretation of nuclear safety advocates a more comprehensive view of the activities that support the more ‘aims and content of nuclear safety’. This multidisciplinary approach is necessary for effective and sustainable nuclear security framework, both at the national and supranational levels. Generally speaking, nuclear safety can be defined as a set of measures to ensure the conditions for the operation of the nuclear facility to prevent nuclear accidents or mitigate their consequences and reduce the radiation risk to the prescribed level. It is well known that nuclear safety cannot be treated outside of the context of responsibility for carrying out nuclear activities and nuclear damage caused by a nuclear accident, and thus the insurance to cover liability for nuclear damage. [ABSTRACT FROM AUTHOR]

Publics and policymakers increasingly have to contend with the risks of complex, safety-critical technologies, such as airframes and reactors. As such, 'technological risk' has become an important object of modern governance, with state regulators as core agents, and 'reliability assessment' as the most essential metric. The Science and Technology Studies (STS) literature casts doubt on whether or not we should place our faith in these assessments because predictively calculating the ultra-high reliability required of such systems poses seemingly insurmountable epistemological problems. This paper argues that these misgivings are warranted in the nuclear sphere, despite evidence from the aviation sphere suggesting that such calculations can be accurate. It explains why regulatory calculations that predict the reliability of new airframes cannot work in principle, and then it explains why those calculations work in practice. It then builds on this explanation to argue that the means by which engineers manage reliability in aviation is highly domain-specific, and to suggest how a more nuanced understanding of jetliners could inform debates about nuclear energy. [ABSTRACT FROM AUTHOR]

The article focuses on electricity generation through nuclear power in Nigeria. Topics discussed include analysis of the state of nuclear power generation in Nigeria, subsidies provided by government nuclear power and policies related to financing and tax incentives and waste storage, and establishment of the Nigeria Nuclear Regulatory Authority to regulate nuclear safety and radiological protection.

The Fluoride-salt-cooled High temperature Reactors (FHRs) are an advanced concept using a novel combination of high-temperature coated-particle fuel, low-pressure fluoride-salt coolant and air-Brayton power conversion system. RELAP5, developed for the best-estimate transient simulation of light water reactors, is still recommended to perform safety analysis of FHRs during postulated accidents. Thermo-physical properties and transport properties of liquid and vapor fluoride salts are implemented into the RELAP5 MOD3.2 source code, through which the application of the RELAP5 MOD3.2 can be extended from the water-cooled reactors to the FHRs. Specific heat transfer correlations are also incorporated into the code for this type of reactors. A RELAP5 model is developed for the MK1 Pebble Bed Fluoride-Salt-Cooled High-Temperature Reactor (MK1 PB-FHR). Using the modified code, two basic postulated transients including the unprotected loss of forced circulation (ULOFC) and unprotected transient overpower (UTOP), are investigated to explore the safety characteristics of the MK1 PB-FHR. Results show that the core outlet temperature may exceed its limit in the ULOFC caused by two main pumps trip. It can be also concluded that the passive residual heat removal system can mitigate the consequences of the ULOFC. Results of the sensitivity analysis of the UTOP indicate that the reactor power and core temperatures show a linear increment trend versus the positive reactivity insertion. [ABSTRACT FROM AUTHOR]

Sustainable and safe energy supply with High Temperature Gas-cooled Reactor (HTGR) fueled by uranium from seawater have been investigated and discussed. From the view point of safety feature of self-regulation with thermal reactor of HTGR, the uranium resources should be inexhaustible. The seawater uranium is expected to be alternative resources to conventional resources because it exists so much in seawater as a solute. It is said that 4.5 billion tons of uranium is dissolved in the seawater, which corresponds to a consumption of approximately 72 thousand years. Moreover, a thousand times of the amount of 4.5 trillion tU of uranium, which corresponds to the consumption of 72 million years, also is included in the rock on the surface of the sea floor, and that is also recoverable as seawater uranium because uranium in seawater is in an equilibrium state with that. In other words, the uranium from seawater is almost inexhaustible natural resource. However, the recovery cost with current technology is still expensive compared with that of conventional uranium. Then, we assessed the effect of increase in uranium purchase cost on the entire electricity generation cost. In this study, the economy of electricity generation of cost of a commercial HTGR was evaluated with conventional uranium and seawater uranium. Compared with ordinary LWR using conventional uranium, HTGR can generate electricity cheaply because of small volume of simple direct gas turbine system compared with water and steam systems of LWR, rationalization by modularizing, and high thermal efficiency, even if fueled by seawater uranium. It is concluded that the HTGR fueled by seawater uranium with the current technology enables the energy sustainability for approximately 72 million years with superior inherent safety features and lower cost of 6.07 cents/kWh (7.28 yen/kWh) than the 7.34 cents/kWh (8.80 yen/kWh) cost of LWR using conventional uranium. [ABSTRACT FROM AUTHOR]

The nuclear community frequently refers to the concept of 'future generations' when discussing the management of high-level radioactive waste. However, this notion is generally not defined. In this context, we have to assume a wide definition of the concept of future generations, conceived as people who will live after the contemporary people are dead. This definition embraces thus each generation following ours, without any restriction in time. The aim of this paper is to show that, in the debate about nuclear waste, this broad notion should be further specified and to clarify the related implications for nuclear waste management policies. Therefore, we provide an ethical analysis of different management strategies for high-level waste in the light of two principles, protection of future generations-based on safety and security-and respect for their choice. This analysis shows that high-level waste management options have different ethical impacts across future generations, depending on whether the memory of the waste and its location is lost, or not. We suggest taking this distinction into account by introducing the notions of 'close future generations' and 'remote future generations', which has important implications on nuclear waste management policies insofar as it stresses that a retrievable disposal has fewer benefits than usually assumed. [ABSTRACT FROM AUTHOR]

The possibility of applying isotopically modified molybdenum like the structural material of fuel elements for light-water and fast reactors with dispersed fuel containing granules of the metal U-Mo-alloy in Mo matrix, has been discussed. It has been shown that with isotope-modified molybdenum with the neutron capture cross-section close to natural zirconium, the improved safety of nuclear reactors, i.e., both thermal and fast-neutron reactors, due to the improvement of heat-transfer emission in fuel elements has been achieved. Assessments demonstrated that molybdenum with economic practical cost may be obtained on the existing cascades of gas centrifuges developed for the separation of nonuranium isotopes. [ABSTRACT FROM AUTHOR]

In this study, we encompass different designs of small reactors, making a review since the beginning of the concept in the military and academic research, to the actual designs; and the prospective of this kind of reactors in the development of the nuclear industry. Small reactors were developed many years ago and have been in operation since 1950, but now, there is an increasing interest in developing small reactors for different applications or co-generation including a large variety of reactor designs with features ranging in coolant, cycle type, fuel, and energy applications; and enhancing features such as nuclear inherent safety, passive safety, and proliferation resistance. The modularity of some of these reactors makes them also a possible choice for nuclear development of some countries with units manufactured off-site and possible scaling into multimodular plants. Different lines of small reactors and current status are reviewed in this work. [ABSTRACT FROM AUTHOR]

Precision (benchmark) experiments on uranium systems with different moderators performed at the National Research Center Kurchatov Institute in a wide range of enrichment, ratios between the nuclear concentration of the moderator and uranium, as well as temperature are reviewed. The requirements of the arrangement of the experiments and participation of Russian specialists in international projects for the evaluation of the benchmark-experiments in terms of the critical nuclear safety (ICSBEP) and evaluated reactor-physical benchmark-experiments (IRPhEP) (NEA OECD) are discussed. The evolution of the experimental base required for performing such experiments and its current status are briefl y discussed. [ABSTRACT FROM AUTHOR]

For a long time, the domestic system regulating the safety of the use of atomic energy was oriented toward its own structure of the relevant documents. The orientation on synchronization and harmonization of the requirements, characteristic for the last 30 years, recently started to emphasize safety at the final stages of the life cycle. In the present article, on the basis of the IAEA documents as well as the recommendations of the IAEA standards, the main anticipated problems of validating the long-time safety of disposal sites for radwastes are fixed. The main avenues are shown for the control and regulation agencies as well as the scientific-technical community for bringing the practice of validation of long-time safety into conformance with international principles and requirements of the domestic legislation. [ABSTRACT FROM AUTHOR]

The factors supporting the safety of atomic energy are discussed. It is underscored that all factors considered in the generally accepted nuclear-safety strategy must be fully taken into account and implemented in order to achieve this goal. [ABSTRACT FROM AUTHOR]