Accident Management Research Articles

Numerical simulation technologies for safety evaluation in plant lifecycle optimization method, ARKADIA for advanced reactors

This paper describes the development of Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA) which provides best possible solutions for the design, safety measures, maintenance, and decommissioning of advanced nuclear reactors. ARKADIA consists of state-of-the-art numerical simulation methods, the artificial intelligence technologies, and a knowledge base that includes experiences from the development and operation of the fast reactors and the related research and development. Two subsystems of ARKADIA: ARKADIA-Design and ARKADIA-Safety, are currently being developed individually as the first target of sodium-cooled fast reactors. In a subsequent development phase, the two subsystems will be integrated into a single system, ARKADIA, that is applicable to advanced reactors with a variety of concepts, coolants, configurations, and output levels. ARKADIA-Design offers the functions to support design optimization both in normal operating conditions and design basis events mainly during a conceptual design stage in the fields of core design, plant structure design including thermal-hydraulics analysis, and maintenance plan optimization. The main technology of ARKADIA-Design is based on the multi-level simulation approach by the coupled analysis with numerical analysis codes according to the user’s requirements. For instance, the coupled simulation of neutronics, core deformation, and core thermal hydraulics could evaluate the negative feedback effect of the core deformation reactivity during the plant transient. ARKADIA-Safety is developed for the automatic optimization of severe accident management and their feedback to a plant design. ARKADIA-Safety includes the thermal-hydraulics and multi-physics code, SPECTRA, for in- and ex-vessel integrated severe accident simulation. The basic capability to evaluate severe accident progress was demonstrated through the analysis of a hypothetical loss of reactor level event in a sodium-cooled fast reactor.

Measurement of the water-vapor void fraction in a 4x4 unheated rod bundle

ABSTRACT In recent years, computational fluid dynamics (CFD) simulations have been used to evaluate the two-phase flow behavior inside fuel bundles for nuclear core design and accident management. Space – time distributions of the void fraction and interfacial velocity in bundle systems at high temperatures and pressures are important for validating CFD codes describing two-phase flow. However, such space – time distributions of the void fraction and interfacial velocity in bundle systems are difficult to obtain under high-temperature and high-pressure conditions. Thus far, we have developed an experimental apparatus with a 4 × 4 unheated rod bundle by adapting a through-rod wire mesh sensor to measure the void fraction and interfacial velocity distributions under high pressures and high temperatures. In this study, we measured distributions of the void fraction and interfacial velocity in a water–vapor system under high-pressure conditions up to 2.6 MPa and high temperatures, using the developed apparatus. We demonstrate that the experimental apparatus yields reasonable results for two-phase flow inside a fuel bundle. However, the analysis method used in this study could not properly estimate the interfacial velocity under some experimental conditions, highlighting the need to improve the method for analysis of the interfacial velocity.

Use of supporting software tool for decision-making during low-probability severe accident management at nuclear power plants

In the project NARSIS – New Approach to Reactor Safety ImprovementS – possible advances in safety assessment of nuclear power plants (NPPs) were considered, which also included possible improvements in the field of management of low probability accident scenarios. As a part of it, a supporting software tool for making decisions under severe accident management was developed.
 The mentioned tool, named Severa, is a prototype demonstration-level decision supporting system, aimed for the use by the technical support center (TSC) while managing a severe accident, or for the training purposes. Severa interprets, stores and monitors key physical measurements during accident sequence progression. It assesses the current state of physical barriers: core, reactor coolant system, reactor pressure vessel and containment. The tool gives predictions regarding accident progression in the case that no action is taken by the TSC. It provides a list of possible recovery strategies and courses of action. The applicability and feasibility of possible action courses in the given situation are addressed. For each action course, Severa assesses consequences in terms of probability of the containment failure and estimated time window for failure. At the end, Severa evaluates and ranks the feasible actions, providing recommendations for the TSC. The verification and validation of Severa has been performed in the project and is also described in this paper. Although largely simplified in its current state, Severa successfully demonstrated its potential for supporting accident management and pointed toward the next steps needed with regard to further advancements in this field.

The Strategic Communication Planning in an emergency at the Almirante Álvaro Alberto Nuclear Power Plant (CNAAA)

In a world where information has a high speed of propagation and is present in the most diverse media, it is essential to follow the news in order to keep the population informed about the actions taken, the real risks and the consequences of an alleged nuclear accident. Effective public communication is understood to encourage the implementation of appropriate protective actions by people at risk and reassure individuals who are not directly at risk, reducing rumors and fears. Thus, the Strategic Communication Planning (SCP) acquires high relevance in a context of panic control, disseminating the correct procedures that will directly contribute to the coordination of the nuclear emergency. In this sense, this work aims to present the steps of a SCP in the case of a nuclear emergency at the Almirante Álvaro Alberto Nuclear Power Plant. In this case, the SCP is divided into the Risk Management and Accident Management phases, being prepared in order to unite the legislation in force in the country with international agreements, technical standards and civil defense doctrines. It is understood that the SCP makes it possible to efficiently build communication channels with the public, as well as to provide authorities with strategies to disseminate information, creating performance indices to evaluate these strategies in relation to the stakeholders in the communication process, because only with a Implemented SCP will make it possible to create a continuous improvement in the information feedback processes with an increase in the quality of communications

OPTIMIZATION OF ROAD SAFETY PARTNERSHIP ACTION (RSPA) IN HANDLING TRAFFIC PROBLEMS IN THE JURISDICTION OF THE TANJUNG PRIOK PORT POLICE STATION

The Tanjung Priok Port in Indonesia faces challenges in traffic management and road safety. To address these issues, the Tanjung Priok Port Police Traffic Unit has implemented Road Safety Partnership Action (RSPA) activities. This study aims to determine the implementation of Road Safety Partnership Action (RSPA) activities carried out by the Tanjung Priok Port Police Traffic Unit in handling traffic problems in the Tanjung Priok Port area. The study incorporates various theories, including Rosen's cooperation theory, Terry's management theory (including planning, organizing, implementing, and supervising), Dr. E Mulyana's competency theory, and problem-solving theory. A qualitative approach with descriptive analysis was employed. The findings revealed suboptimal implementation of the RSPA activities by the Tanjung Priok Port Police Traffic Unit in addressing traffic and road transportation issues. Furthermore, the competence of the Satlantas Polres Pelabuhan Tanjung Priok personnel in executing the RSPA activities was also deemed suboptimal. The active system and methodology employed in the RSPA activities have not been fully maximized. Therefore, optimization measures are required, including strengthening coordination and communication capabilities, empowering budget support and improving facilities and infrastructure in RSPA activities, enhancing the enforcement of traffic violations and management of traffic accidents, fostering personnel quality in executing the RSPA activities, implementing traffic engineering measures in anticipation of increased port activities related to transportation, encouraging community participation, and utilizing advancements in information technology, particularly in traffic engineering within the Tanjung Priok Port area.

PKP-PK Unit Personnel Strategy Analysis in Mitigating Aircraft Accidents at Adi Soemarmo Boyolali International Airport

This study aims to find out how well and how hard the Aviation Accident Assistance and Fire Fighting (PKP-PK) personnel handle fires at Adi Soemarmo Boyolali International Airport. This research will help by facilitating PKP-PK to follow the safety SOP (Standard Operation Procedure) if an incident occurs. Case study method descriptive qualitative research was used for this study. The PKP-PK unit at Adi Soemarmo Boyolali International Airport was used as a research location. The results of the review show how the progress and progress of the Avionics Accident Management and Fire Fighting (PKP-PK) workforce in tackling fire incidents at the Adi Soemarmo Boyolali World Air Terminal. PKP-PK personnel always carry out their responsibilities in accordance with applicable regulations.

Accident/RTO/Abandoned Vehicle Investigation Using QR Code on Vehicle

Abstract: In India fatal road traffic accidents constitute a significant public health problem in many countries around the world. In Saudi Arabia, road accidents are one of the leading causes of death. The delay in detecting and caring for people who have been exposed to traffic accidents increases the risk of injuries. Injury care after the accident is severely affected by time; a few minutes may separate life from death. Several solutions have been proposed to reduce the incidence of accidents or to reduce the mortality rate at a high financial cost. Recent approaches are using built-in vehicle accident notification system. While these approaches work in pleasing manner, they are expensive, maintenance complex task, and are not available in all cars. However, there is a lack of methods and solutions used to detect the rapid occurrence of accidents and thus speed in transfer of patients to medical centres. In this project we are providing solution for betterment of traffic accident management The various uses we are providing are vehicle identification, in case of accident to identify the victim in case of vehicle stolen-using chasse number, etc. we are using QR code to make the users documents secure. QR Codes have fast response time and have large storage capacity, QR Codes can be used perfectly to send encrypted data (messages) to the receiver. In user panel, user will upload documents after submission the verification takes place in verification panel then the verified user will be given the unique QR code. At the signal, traffic police will scan the QR code and get all the documents of the respective user. So this system will provide a smart and easy way of verification of documents and identifying the user in accident management system

Comparison of random forest and support vector machine regression models for forecasting road accidents

One of the world’s concerns today is the rate of road traffic accidents (RTA). The overwhelming majority of these accidents occur in low and middle-income countries. RTA are one of the leading causes of death in Rwanda. RTA prediction is crucial for both transportation management and intelligent transportation systems (ITS) development. This paper adopted the use of two modelling techniques, Random forest (RF) and Support vector machine (SVM) for short-term road accident forecasting. The data used to evaluate the models was obtained from the Police. The lower error indices of MAE, MSE, RMSE and higher coefficient of determination (R2) were accuracy measures in comparing the models. The RF model performed better than the SVM model as it revealed higher R2=0.91 compared to the SVM model that was with R2=0.86. Machine learning methods are promising tools for the prediction of road accidents. Prediction positively influences safety enhancements and regulation formulation to prevent future accidents. The appropriate prediction method would help policymakers and healthcare providers adjust their contributions to the accident management process.

Collaboration Tripartite Elements Accident Reporting and Potential Hazards Management Coal Mining, Indonesia: Systematic Review

Objective: Occupational accidents and diseases in Indonesia are still a scourge because they not only cause minor and serious accidents, but often result in fatalities. The same occupational accidents and diseases’ would seem to be happening in the coal mining business sector. According to the Badan Penyelenggara Jaminan Sosial Indonesia, 234,370 cases of work accidents and work-related illnesses occurred in 2021, of which the mining sector contributed 6,565 cases. Collaboration between companies, workers and local governments in the management of work accident reporting needs special attention so that similar incidents do not occur in the future.Material and Methods: This literature review supports the researchers hypothesis or belief in disclosing that there is an increasingly open space for collaboration between companies, workers and local governments in accident reporting management to reduce accidents. The literature review search term used is work accident reporting sourced from Google Scholar, National Library of Indonesia, PubMed, and other sources in the 2017-2022 period.Results:This literature review found a number of facts that companies have an obligation as stipulated in the Regulation of the Minister of Manpower of the Republic of Indonesia No. 03 of 1998 concerning Accident Reporting and Investigation Procedures to report every work accident and work-related illness that occurs both periodically and specifically to the competent authority. Likewise, workers have the responsibility to report any accidents and potential hazards found in the workplace to the occupational safety and health officer. The Regional Government is no exception, which is a regulator and at the same time supervises companies in carrying out their business activities.Conclusion: To reduce workplace accidents requires diligent and constructive cooperation between the business managers, the workers, and the government regulators to support a zero-accident work environment.

Development of dynamic integrated consequence evaluation (DICE) for dynamic event tree approaches: Numerical validation for a loss of coolant accident

Although conventional probabilistic safety assessment (PSA) makes a sufficient contribution to evaluating the safety of nuclear facilities, it suffers some limitations in performing optimal assessments close to the actual behavior by reflecting time dependencies in accident management. However, integrated deterministic and probabilistic safety assessment (IDPSA) literally combines two methodologies to enable synchronized evaluation based on physical behavior, system availability of facilities, and operator's response. Dynamic integrated consequence evaluation (DICE) is one of the IDPSA tools developed for this purpose, which is the first achievement in South Korea and supported by a regulatory authority. Based on dynamic discrete event tree (DDET) methodology, DICE consists of a scheduler that supports the exchange of information between modules, including a physical module that computes thermal-hydraulic simulations, a diagnosis module that specifies branching points for safety systems, and a reliability module that provides system availability. This paper deals with numerical validation comparing the conventional deterministic and probabilistic results for a sample scenario with those of DICE, and discusses an application of DICE to support and extend the scope of conventional PSA through a case study. It is expected that DICE emphasizes the compliance with conventional PSA methodologies while providing magnified perception in investigating unforeseen scenarios.

Uncertainty quantification for severe-accident reactor modelling: Set-up and first results of the Horizon-2020 project MUSA

The current Horizon-2020 project on “Management and Uncertainties of Severe Accidents (MUSA)” aims at applying Uncertainty Quantification (UQ) in the modelling of Severe Accidents (SA), particularly in predicting the radiological source term of mitigated and unmitigated accident reactor scenarios. A selected number of severe accident sequences of different nuclear power plant designs (e.g. PWR, VVER, and BWR) are addressed.The application of the Best Estimate Plus Uncertainty (BEPU) methodology to reactor accident scenarios requires a number of key steps: (i) the selection of severe accident sequences for each reactor design; (ii) the development of a reference input model for the specific design and SA-code; (iii) the definition of the figures of merit for the UQ-analysis; (iv) the selection of a list of uncertain model parameters to be investigated; (v) the choice of a statistical tool to propagate input deck uncertainties; (vi) the selection of a feasible approach (i.e., Monte Carlo versus order statistics) to address UQ by using a statistical software (i.e., UQ-tools DAKOTA, SUSA, URANIE, etc.); (vii) the running phase to achieve a high number of successful realizations with the SA codes; and, (viii) the statistical evaluation of the results (i.e., sensitivity analysis).This paper describes each of these steps such as settled in the reactor applications work package of the EU MUSA project and pays particular attention to the choices made by partners. It presents preliminary results also with an emphasis on the major challenges posed by BEPU application in the field of SA analysis.

Improved model for traffic management system using KDD and big data: case study Jordan traffic

This paper addresses the challenge of the increasing number of traffic accidents resulting from population growth and increased vehicle usage, leading to significant economic and environmental impacts. To address this issue, this paper proposes an improved model for traffic accident management that employs Knowledge Discovery in Databases (KDD) and extensive data analysis techniques to extract the main factors contributing to car accidents. A case study of traffic accidents in Jordan is conducted, utilizing actual data from the Department of Statistics and the Public Security Directorate Records between 2016 to 2020 to identify the primary factors contributing to road accidents and their effects on various types of injuries. The study identifies 11 factors that contribute to car accidents, with driver error being the primary factor for increasing the number of accidents and injuries. Additionally, minor car accidents are found to cause more injuries based on the analysis of the proposed approach. The findings provide a solid foundation for developing an accurate and precise scheme to reduce or eliminate the number of road accidents, which is essential for transitioning to smart cities. Using big data and KDD techniques could significantly improve current traffic accident management practices, informing the development of new policies and regulations to improve road safety and reduce economic and environmental impacts. This paper presents a promising solution for addressing traffic accident management, contributing to safer and sustainable urban environments.

Key elements for proper development and implementation of severe accident management guidance/guidelines – SAMG

Severe Accident Management Guidelines (SAMG) have been developed to support a nuclear power plant (NPP) in case of a severe accident, i.e. an (unlikely) accident where the core and/or the fuel in the Spent Fuel Pool (SFP) have been damaged and there is risk for radioactive releases. Various approaches exist, such as the SAMG developed by the PWR and BWR Owners Groups (PWROG, BWROG), by Electricité de France (EdF) for its fleet of plants, the Candu Owners Group (COG) and by several individual NPPs. Often, the developers of SAMG develop a generic type of SAMG, applicable for a certain type of NPP, and leave it to the individual plants to adapt these to their configuration and Emergency Response Organisation (ERO). That means, that quite some burden remains for the individual plants to finally implement the SAMG. This implementation includes not only technical items, such as defining proper strategies, but also many organisational issues, as handling a severe accident needs an effective ERO, in which the SAMG must be well embedded.Various plants have been reviewed, both the ways they have constructed their plant-specific SAMG, implemented them in their ERO and tested the performance in drills and exercises. Most recently, a review of SAMG at Swedish plants has been performed, after which it was decided to pull the various experiences together and summarize them in one compendium. Some of these also include aspects of the development of the generic SAMG.The result is a number of recommendations (∼30), both on the technical field as well as regarding the ERO. We could also define a number of good practices that deserve wider attention.A few words have been devoted to international support for a stricken plant (Epilogue).

Development of a surrogate model for quenching estimation of ex-vessel debris beds and its coupling with MELCOR

In the severe accident management (SAM) strategy for Nordic boiling water reactors (BWRs), a flooded reactor cavity is conceived to receive corium in case of vessel failure, with the hope that the discharged corium will fragment and form a coolable particulate debris bed in the deep water pool. The so-formed debris bed on the cavity basement is supposed to be very hot at the beginning and therefore its quenching is a prerequisite for long-term coolability. In previous study the coupled MELCOR/COCOMO simulation was employed to simulate quench process of ex-vessel debris beds in severe accident scenarios. Although it successfully extended the MELCOR capability, the calculation was dramatically slowed down by explosive computational cost of COCOMO. To overcome the limitation, the present study is to develop a surrogate model (SM) which can replace the mechanical code COCOMO and realize quick estimations of the quench process of ex-vessel debris beds. It was then coupled with MELCOR code for integral severe accident analyses of a Nordic BWR with cooling of ex-vessel debris beds. The SM was developed based on a database generated from COCOMO calculations of various one-dimension (1D) debris beds quenched in the reactor cavity, using artificial neural networks (ANNs). Finally, the coupled MELCOR/SM simulation was applied to safety analyses of postulated severe accident scenarios due to station blackout (SBO) in the BWR, where MELCOR performs integral analysis of accident progression while SM predicts the consequences (e.g. energy transfer) of debris bed quench. The simulation results show that the coupled MELCOR/SM simulation can predict the trends of containment pressure and pool temperature similar to those of the coupled MELCOR/COCOMO simulation. Compared with MELCOR standalone simulation, the coupled MELCOR/SM simulation predicted earlier pool saturation and containment venting.

METHOD OF QUALIFICATION OF PASSIVE SAFETY SYSTEMS OF MODULAR NUCLEAR REACTORS WITH CIRCULATION CIRCUIT FLOWS

Low-power modular reactors are a promising direction for increasing the safety of nuclear power, because accident management in modular reactors is carried out only by passive safety systems (without electric pumps). Critical for the safety of modular reactors are accidents with a violation of the tightness of the natural circulation circuits of passive safety systems. The main limitations of using traditional accident modeling approaches with deterministic codes to qualify the reliability and operability of passive safety systems of modular reactors are related to the possibility of negative effects of "code differences" and "code user differences", as well as the unfoundedness of code verification/validation results. An original qualification method has been developed to ensure the safety conditions of the passive safety systems of the Westinghouse low-power modular reactor (SMR) in the event of accidents with a violation of the tightness of the natural circulation circuits. The assumptions adopted in the developed method ensure the conservatism of qualification results. Based on the preliminary calculation qualification of the natural circulation circuits of the SMR passive safety systems, it was established that for the relative sizes of the leaks, greater than 5% of the pipeline cross-section, a violation of safety conditions and drainage of the active zone may occur less than 24 hours after the start of the accident. It is necessary to modernize the SMR with regard to systems for diagnosing leaks in natural circulation circuits of passive safety systems and isolating damaged sections of circuits.

A scoping study on remelting process of a debris bed in the lower head of reactor pressure vessel

Coolability and retention of core melt (corium) in the lower head of reactor pressure vessel (RPV) has been accepted as a severe accident management strategy to maintain the reactor pressure vessel (RPV) integrity of a light water reactor. To qualify the in-vessel melt retention strategy, lots of studies have been performed to investigate the natural convection heat transfer of a melt pool in the lower head. However, little work has been attributed to the precursory phase of the melt pool, i.e., the remelting process of a debris bed which is formed in the lower head at the very beginning of corium relocation from the core the lower head. The present study is motivated to conduct an experimental study on the debris remelting process. For this purpose, a dedicated test facility named COREM (COrium REMelting) is conceived and constructed, which features internal heating of electromagnetic induction and visualization of debris remelting dynamics. The test section is a semicircular vessel representing a slice of scaled-down RPV lower head, whose front and back walls are made of transparent tempered glass which facilitate visualization and induction heating of the debris bed. Fiber probes with multiple optical temperature sensors are mounted in the semicircular wall of the test section to measure its temperature distribution. In the scoping test, n-octanol and Wood's metal are selected as the simulant materials of metallic and oxidic components of corium, respectively, and their particles ware loaded in the test section to form a debris bed. This paper presents the first two scoping tests which have been performed with the COREM facility so far, under different mixing ratios of two debris materials. The measured data include the photography of debris remelting processes and the temperatures in the debris bed and the semicircular wall. Based on the temperature distributions, heat flux profile along the semicircular vessel wall is also estimated. The scoping tests well reproduce the dynamic process of debris remelting, with two distinct stages of fusion of n-octanol and Wood’s metal, i.e., melting successively from low to high melting-point debris particles. During the remelting process, the vessel wall temperatures increase with the polar angle firstly and then decrease gradually, leading to the highest temperature appearing in the middle of the lower layer of the stratified molten pool which is finally formed.

MELCOR-To-MELCOR Coupling Method in Severe Accident Analysis Involving Core and Spent Fuel Pool

A lot of effort has been spent to prevent the occurrence of SA in nuclear plant and to develop Severe Accidents (SA) Management to mitigate the consequences of a SA. Those consequences are mainly related to limit the release of fission product to the environment. The core in the vessel is not the only source of fission products as the Spent Fuel Pool (SFP) hosting the fuel removed by the core is, in some NPP, inside the containment and SA conditions can also occur. This is especially important in reactors having proximity between the RPV and SFP such as the VVER-1200. This close proximity implies that any SA occurring in the SFP potentially affects the RPV and vice-versa. This potential combination might cause unexpected evolution in the SA progression to whom the safety systems are not able to contain. MELCOR code is a widely used, flexible powerful SA code but it is incapable (due to the uniqueness of the COR package use inside the same input) to reproduce a situation in which both the fuel in vessel core and the fuel in the SFP, inside the same containment, are going to experience a severe accident scenario. The current study presents a MELCOR-to-MELCOR coupling method to simulate simultaneously scenarios with both, core and SFP, as sources capable of H2 generation, fuel damage and FP release in a VVER-1200 NPP. The coupling is performed by running two simulations in parallel and with the data exchange supervised and managed by a dedicated Python coupling supervising script developed at NINE.

Backfitting of Dukovany NPP for design extension conditions

The contribution deals with the backfitting of Dukovany NPP for design extension conditions with emphasis on severe accidents. This NPP, located in Czech Republic, is equipped with 4 units of VVER-440/V213 reactors. During the last decades significant attention was paid to backfitting of these units. New safety systems aimed at mitigating the consequences of severe accidents have been designed and have been (are being) implemented. At first, special features of VVER-440/V213 reactor and containment equipped with bubbler condenser pressure suppression system are briefly described. The systematic process of plant backfitting for design extension condition is summarized further. Attention is paid to description of crucial severe accident management strategies that have been adopted and design and implementation of new safety systems that are necessary to fulfil these strategies. Than the requirements of Czech regulatory body (SUJB) set for practical elimination of early or large radioactive releases are discussed. Finally, the performances of severe accident management strategies are illustrated using the specific case of blackout scenario. For this purpose, an analysis was performed using ASTEC code developed by the French IRSN. The results of analysis are briefly described and discussed further. From the obtained results it follows that the high-pressure melting scenario is reliably avoided, corium is localised and stabilised inside the reactor vessel and the hydrogen in containment atmosphere is safely controlled by passive autocatalytic recombiners. Thus, stable plant state with preserved containment integrity is reached. Source term that has been estimated under these conditions indicates that the predicted releases are well below the criteria set by SUJB for early or large radioactive releases, and therefore long-term contamination of the environment is avoided.

Effects of a Theory-based Educational Intervention on Skills of Emergency Medical Service Personnel for Scene Management in Chemical Accidents

Background: Scene management in accidents reduces losses, speeds treatment, and facilitates immediate victim transfer. Learning the related skills through theory-based educational intervention makes this task possible. This study aimed to determine the effect of theory-based educational intervention on emergency medical personnel skills to manage the scene in chemical accidents. Materials and Methods: The present study employed a quasi-experimental design with pre- and post-intervention measurements. Using the random sampling method, 70 emergency medical technicians working in Qazvin Province, Iran, were selected and assigned to experimental and control groups. The data collection tools included demographic questions and scales to assess the theory of planned behavior. The intervention program consisted of 4 sessions. Data were analyzed by SPSS software, version 23 using descriptive statistics and analysis of covariance. Results: The Mean±SD age of the participants was 35.03±5.57 years, and 82.6% (57 people) were married. After the intervention, the Mean±SD scores of attitudes (40.11±4.77), subjective norms (15.06±4.27), perceived behavior control (18.53±2.55), and behavioral intention (8.49±2.27) increased significantly in the experimental group (P<0.001). Also, the self-report behavior in the experimental group improved significantly (P<0.001). Conclusion: Theory-based educational intervention improved psychological variables and safe behaviors in chemical accident management. Implementing educational interventions based on this theory enhances the safety performance of medical emergency technicians in chemical accidents.

Unusual Intracaecal Migration of a Intrauterine Device: The Place of Laparoscopy in the Diagnosis and Treatment

The intrauterine device (IUD) is an effective and widespread contraceptive method. Its transuterine migration into the pelvic organs or the digestive tract is a rare and sometimes serious accident. We report the case of a 35-year-old patient who presented with transuterine migration of an intrauterine device placed 2 years earlier. Imaging confirmed migration and suspected the digestive tract localization of the IUD. With laparoscopic approach the intra-cecal localization was precise and the treatment made with good postoperative course. Through this observation we emphasize the diagnostic and therapeutic role of laparoscopic approach in the management of this contraceptive accident.