Epistemic uncertainty estimate of human errors during nuclear research reactor emergencies using confidence-boxes

Indonesia has three nuclear research reactors, nemely: RSG-GAS in Serpong, TRIGA2000 reactor in Bandung, and Kartini reactor in Yogyakarta. Besides being useful, nuclear reactor has inharent risk due to its radiation hazard. The research reactor nuclear contingency plan was prepared for nuclear preparedness. The study aimed to analyze the implementation of the research reactors’ nuclear contingency plans. The study used an analytic descriptive qualitative methods. The results showed that the nuclear contingency plans of the three research reactors have different implementation range due to their different hazard risks. However, the nuclear emergency responses were the same, they were: sheltering, evacuation, provision of KI tablets, and prohibition of consumption of local food and contaminated river/ ground water. The research reactors’ nuclear contingency plans have not discussed criteria used for taking nuclear emergency responses. The results also showed that the nuclear contingency plan of RSG-GAS in Serpong and the Kartini reactor in Yogyakarta have been implemented partially. While for the TRIGA2000 reactor nuclear contingency plan in Bandung has not been implemented yet. A legal umbrella is needed to implement the research reactor nuclear contingency plan. Thus, coordination and active role from all stakeholders are needed to be enhanced in order to be able to resolve the existing obstacles and challenges.

Poly(aryl ether ether ketone) (PEEK) is a high‐temperature engineering thermoplastic used in the aerospace industry. To assess its suitability for nuclear industry applications, the effects of mixed field radiation from a nuclear reactor were investigated on two industrial semicrystalline PEEK grades (VICTREX 150P and 450P). Specimens were first processed on an ENGEL55 injection molder and then irradiated in the pool of a SLOWPOKE‐2 nuclear research reactor. Specimens were thus exposed to a mixed field of radiation including gamma, electrons, protons, and neutrons, for exposures resulting in doses ranging from 0.15 to 15 MGy. Irradiated samples were characterized by viscosity measurements in the melt, differential scanning calorimetry (DSC), and density measurements. Melt viscosity of irradiated samples proved to be significant in detecting early molecular weight changes in PEEK and helped explain the interaction of competing chain scission and crosslinking reactions at various stages of the radiolysis. All results pointed toward the following transformation of the resin with increased dosage: chain scission of the tie‐molecules between the amorphous and crystalline phases, crosslinking in the amorphous phase, followed by a general degradation of the molecular weight dominated by chain‐scission. © 2002 Wiley Periodicals, Inc. J Appl Polym Sci 86: 2713–2719, 2002

Disastrous circumstances such as radiation and nuclear meltdown require a large and complex scale of emergency health and social care capacity planning framework. Devastating issues and challenges of incompleteness, inconsistency, and the infeasibility of provided requirements of the suggested planning framework might create unnecessary conflicts during the system development. In this paper, we proposed requirements engineering of an emergency preparedness and response simulation model (EPRM) for Malaysia Radiation and Nuclear (MRN) emergency plan simulator. This simulator is expected to plan and manage emergencies and disasters included the risk involved when the nuclear reactor is malfunctioning due to natural disasters such as floods, earthquake, and cyber-attacks. We identify the social and technical problems of implementing the emergencies and disasters policies and also the critical and safety of software systems use to run the nuclear plants. We simulate the proposed emergency preparedness and response model (EPRM) through the simulation software. Finally, we measure the effectiveness of the proposed model using thematic synthesis and qualitative regression analysis provided outstandingly sociotechnical aspects of the affected junctures under investigation. From the testing of the proposed model, we found that if the organization is not able to define and identify, the disaster coordinator roles and responsibility, resources and equipment may contribute 65.63% of emergency plan disorder and severe calamities towards first responders, operators, workers, patients and community at large. Also, we found that those themes were to bone up the main simulation workflow diagram. This finding supported and correlated independently towards Information Systems Development (ISD) key concepts by 13.30 (T value). Those themes were confirmed and reliable coherently (average 89% degree) rather than by chance. Finally, the proposed emergency preparedness and response model (EPRM) that is the theory building methods integrated technical and analytical procedures in a sociotechnical outlook. These methods established on developing emergency response perquisites rather than intervention principles alone. Most likely, this approach was significantly useful to justify a mixture of tacit and explicit knowledge among the emergency plan expertise. Consequently, a strategized, simplified and prevailing RANEP simulator can be achieved, though it is certainly as complex.

The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the Triga IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors.

The work presented in this paper characterize the fiber Bragg grating (FBG) for temperature measurement for the safe operation in material testing reactors (MTR). Fibre Bragg Grating is the key technology used to covers many applications in nuclear reactors in order to sense different physical parameters such as temperature, strain, etc. This paper investigates the modelling of Bragg grating based on MATLAB, through solved coupled mod equations in order to achieve maximum reflectivity. In addition, modelling of fiber Bragg grating as a temperature sensor for nuclear applications using opti-grating software is presented. Furthermore, Apodization techniques for signal processing also used to provide accurate measurements which required in nuclear applications.

Present work discusses the issues of creating the instrumentation for testing the semiconductor magnetic field sensors during their irradiation with neutrons in nuclear reactors up to fluences similar to neutron fluences in steady-state sensor locations in ITER. The novelty of the work consists in Hall sensor parameters being investigated: first, directly during the irradiation (in real time), and, second, at high irradiation levels (fast neutron fluence > 1018n/cm2). Developed instrumentation has been successfully tested and applied in the research experiments on radiation stability of magnetic sensors in IBR-2 (JINR, Dubna) and VVR-M (PNPI, Saint-Petersburg) reactors.

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

Power stabilization is a critical issue in nuclear reactors. A conventional PD controller is currently used in Egypt's second testing research reactor (ETRR-2). This controller is designed neglecting the nonlinear nature of the plant. In this paper, two fuzzy controllers are used instead of the conventional PD. The first fuzzy controller has a static rule set, while the second one has an adaptive rule set. Simulation results show that the adaptive fuzzy controller gives the best integral square error (ISE) index, while the static fuzzy controller still provides a better ISE than the conventional PD controller.

Indian nuclear research reactors provide facilities to carry out irradiation experiments on the core structural materials to assess the radiation damage in these materials. The radiation damage in the material strongly depends on the temperature of irradiation. The temperature-controlled instrumented irradiation capsule (TCIIC) is a device which facilitates irradiation of specimens at a constant higher temperature in a reactor. This paper discusses design of PID controller to establish a constant temperature in the TCIIC system. The control parameters are estimated through placement of dominant pole-zero in the “S-plane” and a stable region of the PID controller parameters was defined. The time response result from the proposed method has been compared with other well-known PID controller design methods and also verified experimentally. The proposed method shows that the controller parameters can provide a stable and reliable performance of temperature control needed in the irradiation experiment.

The recent development status of nuclear power technology in Indonesia includes the operation of three research reactors: GA- Siwabessy, Kartini Yogyakarta, and Triga Mark II Bandung, which have been operating for 31, 39, and 54 years, respectively. With over 30 years of nuclear research reactor operation, one of the issues that come to fore is ageing management. Ageing management includes the duties and issues involved with management, maintenance, operation, monitoring, and refurbishment. Nuclear engineers, researchers, and officials working at the Indonesia Nuclear Energy Agency and Indonesia Nuclear Energy Regulatory Agency have concerns and differing opinions on how these needed functions should be performed and implemented. Use of International Atomic Energy Agency technical documents is beneficial for many reasons, including the global basis and years of experience, however there are a variety of ageing management related documents that it is difficult to discern which ones are applicable to nuclear research reactor programs. This paper attempts to identify, clarify, and filter necessary ageing management documents to be used in the Indonesian nuclear research reactor program.

Background: Worldwide concerns regarding health effects after the Chernobyl and Fukushima nuclear power plant accidents indicate a clear need to identify short- and long-term health impacts that might result from accidents in the future. Fundamental to addressing this problem are reliable and accurate radiation dose estimates for the affected populations. The available guidance for activities following nuclear accidents is limited with regard to strategies for dose assessment in health risk studies.Objectives: Here we propose a comprehensive systematic approach to estimating radiation doses for the evaluation of health risks resulting from a nuclear power plant accident, reflected in a set of seven guidelines.Discussion: Four major nuclear reactor accidents have occurred during the history of nuclear power production. The circumstances leading to these accidents were varied, as were the magnitude of the releases of radioactive materials, the pathways by which persons were exposed, the data collected afterward, and the lifestyle factors and dietary consumption that played an important role in the associated radiation exposure of the affected populations. Accidents involving nuclear reactors may occur in the future under a variety of conditions. The guidelines we recommend here are intended to facilitate obtaining reliable dose estimations for a range of different exposure conditions. We recognize that full implementation of the proposed approach may not always be feasible because of other priorities during the nuclear accident emergency and because of limited resources in manpower and equipment.Conclusions: The proposed approach can serve as a basis to optimize the value of radiation dose reconstruction following a nuclear reactor accident.Citation: Bouville A, Linet MS, Hatch M, Mabuchi K, Simon SL. 2014. Guidelines for exposure assessment in health risk studies following a nuclear reactor accident. Environ Health Perspect 122:1–5; http://dx.doi.org/10.1289/ehp.1307120

TRIGA Bandung is a research nuclear reactor owned by Indonesia, located in Bandung with a power of 2 MWth. Nuclear research reactor TRIGA Bandung is used as a center for applied research and development in the field of application of the nuclear technologies. TRIGA Bandung is currently still using a cylindrical fuel element, this raises a new problem - the limited number of existing fuel element. The purpose of this research is the development of the preliminary core design of a nuclear research reactor TRIGA Bandung using fuel element plate MTR. The research method is modeling and simulation the preliminary design core of nuclear research reactor TRIGA Bandung using comparative method of porous media and non-porous media with CFD code. This research shows the velocity flow and temperature distribution and the influence of pressure from the comparison method of k-ε standard model and porous media model at the preliminary design in the core area of TRIGA Bandung research reactor with fuel element plate MTR.

In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing conditions, such as a control-rod movement, induced by an operator, or an accident scenario. The current methods for transient analysis apply deterministic solvers to calculate the neutronic response, but not only do these deterministic solvers always need to discretise the problem, they also often apply more fundamental approximations such as homogenization or diffusion. Therefore a stochastic method is needed, which only has a statistical uncertainty. A challenge for applying the Monte Carlo method on transient analysis, is the different types of particles which must be simulated in a single simulation. It is common for a neutronics calculation to simulate the prompt neutrons, however the simulation of precursors is new. The main challenge when simulating precursors lies in their long lifetime, which is seconds, whereas the prompt neutrons have a lifetime of microseconds. This has been solved by dividing the transient problem into time intervals and forcing the precursors to produce a delayed neutron is all time intervals. This will ensure prompt-neutron chains in all intervals. The prompt neutrons will form prompt-neutron chains, which poses another challenge for the Monte Carlo method. The chain length of these prompt neutrons varies a lot and the can also split into many branches, which averages out in a real power reactor, but is difficult to simulate on a computer. To improve the chain-length statistics, a new variance reduction technique is developed, which dictates that a single neutron will emerge from each interaction. This emerging neutron can be the result of a scattering event or a fission event. Finally, a dynamic simulation scheme, which can run in parallel, has been devised, simulating all time intervals consecutively. Also, a method for sampling the initial conditions is created and this scheme is implemented in a purpose-built Monte Carlo code and in a general-purpose Monte Carlo code. The two codes have been tested in several cases and they behave as expected, agreeing with deterministic method where expected, but deviating when the deterministic methods are no longer valid. In order to simulate realistic transients in a power reactor, feedback has to be taken into account. To achieve this, the dynamic Monte Carlo method has been coupled to a thermal-hydraulics code, using an explicit scheme. The results of this coupled simulation are compared to a state-of-the-art coupled diffusion calculation in a NURISP-benchmark calculation and the results agree well, except for a small deviation towards the end. The exact source of this deviation should be further investigated, but the difference is small, when realising that there is already a deviation in the steady-state results. The results demonstrate the feasibility of performing a fully dynamic transient analysis, using only Monte Carlo for the neutronics part of the calculation. Transients can now be calculated, with detailed modelling of any complex geometry and with continuous energy, which is especially useful for newly developed reactor types and one-of-a-kind research reactors.

This annual report discusses environmental monitoring at two manufacturing and test sites operated in the Los Angeles area by Rocketdyne Propulsion and Power of Boeing North American, Inc. These are identified as Area 4 of the SSFL and the De Soto site. These sites have been used for research and development (R and D), engineering, and testing in a broad range of technical fields primarily in energy research and nuclear reactor technology. The De Soto site had research and development laboratories involved with nuclear research. This work was terminated in 1995 and only D and D activities will have potential for impact on the environment. Since 1956, Area 4 has been used for work with nuclear materials, including fabricating nuclear reactor fuels, testing nuclear reactors, and dissembling used fuel elements. This work ended in 1988 and subsequent efforts have been directed toward decommissioning and decontamination of the former nuclear facilities. The primary purpose of this report is to present information on environmental and effluent monitoring of DOE-sponsored activities to the regulatory agencies responsible for oversight. Information presented here concentrates on Area 4 at SSFL, which is the only area at SSFL where DOE operations were performed.

In-pile testing of ITER first wall mock-ups at relevant thermal loading conditions in the LVR-15 nuclear research reactor