Accident Initiation Research Articles

Total Loss of Feedwater Analysis of Pressurized Water Reactor Using RELAP5

Abstract In Europe, the design extension conditions (DEC) were introduced after the Fukushima Dai-ichi accident as preferred method for giving due consideration to the complex sequences and severe accidents without including them in the design basis conditions. The objective of the study is to determine available elapsed time before core uncovery and needed DEC safety features for total loss of all feedwater (TLOFW) in a two-loop pressurized water reactor. RELAP5/MOD3.3 computer code has been used for calculations. The initiating event for TLOFW is multiple failures in which, in addition to the loss of main feedwater, the auxiliary feedwater is also lost. The scenarios without DEC safety features and the scenarios with DEC safety feature assumed have been simulated. The results showed that after TLOFW event initiation, it is very important to trip the reactor as soon as possible. In case of loss of offsite power, the reactor coolant pumps stop, and the reactor very quickly trips on low reactor coolant pump flow. When normal operation systems are assumed, the reactor trip occurs on low-low steam generator narrow level few tens of seconds after accident initiation, resulting in less time available before core uncovery occurence. The results for TLOFW scenarios with normal operation systems and DEC safety features assumed demonstrated that secondary side bleed and feed can prevent core uncovery in case when no operator actions are credited in the first 30 min of event. When primary side bleed and feed is used, less time is available for operator actions.

SIMMER modelling of accident initiation phase in sodium fast reactors

The SIMMER-III code was developed mainly for simulation of hypothetical reactor accidents after core melting, but can be applied also for accident initiation phase simulations in sodium and other liquid–metal-cooled fast reactors. New thermal hydraulic and neutronic simulation approaches and models have been developed for treatment of the initiation phase. Two examples of simulation of unprotected loss of coolant flow (ULOF) transients, in ESFR-SMART and FFTF reactors, are presented. It can be concluded that the SIMMER code has a large potential for application to initiation phase analyses.

Dose Calculation for Emergency Control Room HVAC Filter

NPP Krsko is introducing Emergency Control Room (ECR) as part of safety upgrades. According to 10CFR50 Appendix A, GDC 19, both main control room and emergency control room should have adequate radiation protection to permit operators to shutdown the plant and keep it in safe shutdown conditions without receiving more than 50 mSv effective whole body dose, within 30 days from accident initiation. One of the important prerequisites to achieve that is proper operation of control room HVAC. In this work we are focused to calculation of gamma doses from radioactive materials accumulated in HEPA and charcoal filters during 30 days of HVAC operation. The dose at selected points around the filter was calculated using Microshield 10.0 point kernel code. The radioactive gamma source is calculated using RADTRAD 3.03 for plant's severe accident SGTR sequence calculated with MAAP 4.0.7 code. Calculated dose rates at peak filter activity are compared against results obtained with SCALE 6.2 MAVRIC shielding sequence (Monaco Monte Carlo functional module and CADIS methodology). The reasonable agreement between point kernel and hybrid Monte Carlo results was obtained.

Preliminary design of safety system using phase change material for passively cooling of nuclear reactor containment building

The new application of phase change material for passively cooling the containment building of a nuclear power plant provides many benefits such as minimized wall penetration, low maintenance cost, and the possibility of installation in currently operating power plants with a minimal design change. In this study, the applicability of PCM as a passive containment cooling system is verified through design and evaluation of the PCM condenser. In addition, the effective heat capacity method is verified for simulating the melting process of PCM in the condition of a reactor accident. The PCM condenser design is optimized with five design parameters; melting point, volume, heat transfer area, effective thermal conductivity, and total heat absorbed. The performance of the PCM condenser is evaluated by containment safety analysis and is compared to that of an active containment cooling system. Although the PCM condenser had a limitation for absorbing only a fixed amount of thermal energy due to the absence of an external heat sink, the PCM condenser shows better performance in the early stage after the accident initiation compared to the single spray system and finally maintains the integrity of the containment building.

Методика оценки риска при прогнозировании последствий аварий на объектах трубопроводного транспорта

The choice of risk assessment procedure is one of the essential stages of efficient structuring of processes on prevention, localization and elimination of the consequences of accidents at main pipeline transport facilities. The authors analyzed themed publications and regulatory documents, governing procedures of risk assessment and forecasting of the consequences of possible accidents, and defined main problems in this area. Procedure for the risk assessment of accidents at main pipeline facilities was developed, the basis of which is the expert evaluation method. The procedure includes the determination of the main criteria for the assessment the probability of accident initiation and development and the evaluation of the severity of its consequences, an expert evaluation of criteria significance, their classification, and creation of a rating for hazardous pipeline sections. The application of the procedure application allows to specify the list of facilities that require high priority forecasting of accidents consequences, thus to optimize the distribution of resources and the overall increase of efficiency in planning while defining forces and special technical devices, necessary for containment and rectification of emergencies. Expert evaluation method application allows considering the specifics of certain enterprises, their technical and technological peculiarities, thereby increasing forecasting accuracy.

A systematic approach to identify initiating events and its relationship to Probabilistic Risk Assessment: Demonstrated on the Molten Salt Reactor Experiment

One of the first steps in developing a risk assessment model is an exhaustive search for initiating events, which is a systematic and comprehensive starting point to answer the question “what can go wrong?” for a given system design. Identifying Postulated Initiating Events (PIEs) for a reactor design that is at a conceptual or preliminary stage facilitates the incorporation of risk insights into the next iteration of the design process and allows for the early establishment of more quantifiable risk assessment models, such as event sequence diagrams and event tree analysis. Liquid-Fueled Molten Salt Reactors (LF-MSRs) are an example of an advanced reactor technology that does not benefit from having a wealth of operating experience or prior risk-informed safety assessment efforts. Furthermore, design details, such as normal operating conditions and the composition of radioactive material inventories, can deviate substantially from those in other reactors, such that a systematic and comprehensive approach to identifying PIEs for an LF-MSR may highlight accident initiators that have not previously been identified. In the present work, the Master Logic Diagram (MLD) and Hazards and Operability (HAZOP) study approaches were used, together, to identify and consider PIEs for multiple inventories of radioactive material across various Plant Operating States (POSs) in a specific LF-MSR design -- the Molten Salt Reactor Experiment (MSRE). Potentially risk-significant PIEs identified during the analyses of the MSRE design are presented. Furthermore, considerations for exhaustively identifying PIEs for advanced reactor designs are discussed; for example, the combination of inductive and deductive methods was found to provide a robust identification of PIEs in a way that is conducive to the analysis of a nuclear reactor design at an early design stage.

Performing a multi-unit level-3 PSA with MACCS

MACCS (MELCOR Accident Consequence Code System), WinMACCS, and MelMACCS now facilitate a multi-unit consequence analysis. MACCS evaluates the consequences of an atmospheric release of radioactive gases and aerosols into the atmosphere and is most commonly used to perform probabilistic safety assessments (PSAs) and related consequence analyses for nuclear power plants (NPPs). WinMACCS is a user-friendly preprocessor for MACCS. MelMACCS extracts source-term information from a MELCOR plot file.The current development can combine an arbitrary number of source terms, representing simultaneous releases from a multi-unit facility, into a single consequence analysis. The development supports different release signatures, fission product inventories, and accident initiation times for each unit. The treatment is completely general except that the model is currently limited to collocated units.A major practical consideration for performing a multi-unit PSA is that a comprehensive treatment for more than two units may involve an intractable number of combinations of source terms. This paper proposes and evaluates an approach for reducing the number of calculations to be tractable, even for sites with eight or ten units. The approximation error introduced by the approach is acceptable and is considerably less than other errors and uncertainties inherent in a Level 3 PSA.

Integral PWR-Type Small Modular Reactor Developmental Status, Design Characteristics and Passive Features: A Review

In recent years, the trend in small modular reactor (SMR) technology development has been towards the water-cooled integral pressurized water reactor (iPWR) type. The innovative and unique characteristics of iPWR-type SMRs provide an enhanced safety margin, and thus offer the potential to expand the use of safe, clean, and reliable nuclear energy to a broad range of energy applications. Currently in the world, there are about eleven (11) iPWR-type SMRs concepts and designs that are in various phases of development: under construction, licensed or in the licensing review process, the development phase, and conceptual design phase. Lack of national and/or internatonal comparative framework for safety in SMR design, as well as the proprietary nature of designs introduces non-uniformity and uncertainties in regulatory review. That said, the major primary reactor coolant system components, such as the steam generator (SG), pressurizer (PRZ), and control rod drive mechanism (CRDM) are integrated within the reactor pressure vessel (RPV) to inherently eliminate or minimize potential accident initiators, such as LB-loss of coolant accidents (LOCAs). This paper presents the design status, innovative features and characteristics of iPWR-type SMRs. We delineate the common technology trends, and highlight the key features of each design. These reactor concepts exploit natural physical laws such as gravity to achieve the safety functions with high level of margin and reliability. In fact, many SMR designs employ passive safety systems (PSS) to meet the evolving stringent regulatory requirements, and the extended consideration for severe accidents. A generic classification of PSS is provided. We constrain our discussion to the decay heat removal system, safety injection system, reactor depressurization system, and containment system. A review and comparative assessment of these passive features in each iPWR-type SMR design is considered, and we underline how it maybe more advantageous to employ passive systems in SMRs in contrast to conventional reactor designs.

Assessment of emergency consequences and hazardous factors impact from the process of explosive conversion products and materials disposal

Purpose is development of tools for assessing the emergency hazard of facility for disposal of explosive conversion products and materials. Results have been obtained by means of the economic and mathematical modelling methods, with use of the fundamental provisions of probability theory. The main task of conversion products and materials disposal is the creation and organization of safe technological processes aimed at returning the material resources contained in the conversion products to the state’s economy after the appropriate processing, rather than at destroying these resources. As a methodological basis for assessing the emergency hazard of a facility, a provision has been accepted that an integral measure of the disposal process hazard is the economic assessment of accidents. A scheme has been constructed of successiveness of the disposal facility events and states that can lead to an accident, on the basis of which a probabilistic model of its occurrence is developed. Proposed probabilistic model of an accident initiation at the facility for disposal of explosive conversion products and materials makes it possible to predict the behaviour of facility, based on results of the equipment and service personnel state survey, as well as on the analysis of the environmental conditions.

Matrix based methods for assessment of risks induced by complex technical systems

The paper presents a matrix based probabilistic approach to: (i) describing the process of accident initiation and propagation at four scale levels: the level of structural material, the level of individual components of a complex technical system, the systemic level, and the upper level of the system environment, and (ii) developing a risk assessment framework. The proposed approach is based on examining a chronological series of the subsequent discrete failure events occurring at various scale levels of the complex technical systems and its environment. This allows assessing not only the so-called direct risks that take into account for the expected monetary values of direct consequences of the accidents at the system disregarding losses inflicted by subsequent disturbances in the environment of the damaged system, but also indirect risks associated with secondary failures in the system environment. The approach provides an opportunity to describe multilevel fracture in a probabilistic formulation and to obtain an express assessment of risks induced by the complex technical systems.

Implication of LOCA characteristics of large PWR and SMR for future development of intelligent nuclear power plant control system

Operation with reduced number of operators is considered as an important operational mode to improve economics of a nuclear power plant (NPP) in the future and it will be more important for the small modular reactors (SMRs). To control NPPs efficiently and safely with a reduced number of operators, an intelligent operation system can be very useful. However, reduced number of operators will result in larger consequence in human error due to simultaneous multi-unit failures are now more probable than the conventional operational mode. Thus, to decrease the probability of human error occurrence for the reduced number of operators, the development of an autonomous operation system with certain level of intelligence is inevitable. The prediction capability of system’s behavior as well as understanding the root cause for the current situation should be an essential function of the designed autonomous operation system. The authors propose to use the current system thermal–hydraulic (STH) analysis code platform to develop such intelligence. As a demonstration of the suggested method, the authors used the STH code to track the accident initiator of both large pressurized water reactor (PWR) and SMR. The discussions on how this method can be useful for future NPP operation and how generic differences in system design actually influence the system behavior are presented.

Design and Comparative Analysis of Small Modular Reactors for Nuclear Marine Propulsion of a Ship

The fast growth in the size and difficulty of nuclear power plant in the 1970s produced an interest in smaller, modest designs that are intrinsically safe over the usage of design features. With the development of nuclear technology, there is the need for revolution in the Maritime sector, especially the advance marine propulsion. In current years, numerous reactor manufacturers are dynamically improving small modular reactor designs with even superior use of safety features. Several designs integrate the ultimate in greater safety. They totally remove specific accident initiators from the design. Other design features benefit to reduce different types of accident or help to mitigate the accident’s consequences. Although some safety features are mutual to maximum SMR designs, irrespective of the coolant technology, other features are specific to liquid-metal cooled, water, gas, or SMR designs. Results: There have been more reactor concepts investigated in the marine propulsion area by different assemblies and research laboratories than in the power generation field, and much can be learned from their experience for land applications. The extensive use of safety features in SMRs potential to make these power plants extremely vigorous, protecting both the public and the investor. Conclusion: For these two considerations, it is recognized that a nuclear reactor is the ideal engine for naval advanced propulsion. The paper will present the work to analyze the concept design of SMRs and design a modular vessel consisting of a propulsion module.

Preliminary forensic engineering study on aggravation of radioactive releases during the Fukushima Daiichi accident

Even after 6years since the Fukushima accident, the exact accident progression for each unit and location of core debris have not been clarified, although solidified low-melting metal debris was identified at the bottom of the 1F3 reactor pressure vessel (RPV) in July 2017. Currently efforts are directed towards robotic inspection with remote cameras, as well as dose and temperature measurements of the environment inside the Primary Containment Vessels (PCV). In spite of their effort, the observed environmental temperature distribution data does not support the existence of a significant radiation heat sources attributable to the molten core at the bottom of the PCV. At this point the total decay heat of the core debris should be as large asa few hundred kilowatts after 2000days since the initiation of the accident as summarized in Annex A. In addition, the temperature of the accumulated water inside PCVs is 10–20°C even with a reduction of the water injection to 3.0m3/h. The RPVs appear to be holding the heating core debris with water, implying that “in-vessel retention” of core debris has been achieved thanks to effective accident management. This should help greatly in the retrieval of the core debris by removing the top head of RPV.Under these circumstances the author has conducted a forensic engineering study (i.e., different fields of science working together to collect and integrate independent evidences) to clarify the most likely accident scenarios of the Fukushima Daiichi accident. Through this study the author identified that a large portion of the land contamination observed at the north-west direction is mostly the result of the accident that occurred at Unit 2.In this unit its blowout panel, provided for over-pressure protection against a main steam pipe breach in the secondary confinement building, was inadvertently activated before a leakage of radioactive effluent from the PCV. Its activation is believed due to the hydrogen explosion in Unit 1 which occurred on March 12, next day of the Fukushima accident initiation. By losing the confinement function, the radioactive effluent leaked from the 1F2’s PCV and would have been discharged without mitigation. This accident scenario explains the series of leakage events identified in two of the 24 monitoring posts which had been installed by the Fukushima Prefectural Government. A series of six large releases were repeated between March 15 and 16, behaving like a periodical actuation of the safety valves for the PCV. Such multiple release events were very likely induced by the overpressure release of the PCV due to leakage of the dry well flange. This leakage should have been induced through discharging steam and hydrogen due to the activation of Safety and Release Valves (SRV) into the suppression pool (SP) water. Unfortunately the wet well atmosphere must have been that of air, since there was no nitrogen charge line to the atmosphere of the SP surface water. The resultant air-hydrogen mixture resulted in an “internal hydrogen explosion” which should have deformed the flange. The recent robotic inspection inside PCV revealed that a gigantic water splash appears to have occurred at the bottom of the PVC dislodging the gratings installed over the platform.Next to the series of large releases from Unit 2, Unit 1 also induced two large releases on March 12. Fortunately, these releases left more than 3 orders of magnitude less soil contamination compared with the series of releases from Unit 2. Unit 3 also released a significant amount of radioactive species twice on March 13, resulting in a very small soil contamination. The main constituent of the radioactivity is likely radioactive noble gases in this unit.

Experimental estimation of the effect of contact condensation of steam–gas mixture on VVER passive safety systems operation

Results of experimental study of the effects of contact condensation of steam–gas mixture on the operation of VVER passive safety systems and steam generator in emergency condensation mode are presented. Contact condensation takes place when subcooled liquid is supplied in the accumulator tank of VVER reactor auxiliary passive core flooding system (the HA-2 system) in the presence of accumulated non-condensable gases. Water supplied to hydro accumulators can be used for increasing the operating time of VVER steam generator in the emergency condensation mode and for ensuring core cooling during longer time. Low liquid outlet velocity (less than 1m/s) caused by the necessity to ensure safety systems operation in passive mode constitutes the distinguishing feature of the investigated processes.Experiments were conducted on the test facility with parameters typical for the primary cooling loop of the reactor facility 24h after the accident initiation for different concentrations of gas in the steam–gas mixture. Nitrogen and helium, which replaces hydrogen for the purposes of safety assurance, were used as the non-condensable gases. It was established according to the results of experiments that the increase of concentration of non-condensable gas within the volume of HA-2 hydro accumulator model up to 45% leads to the reduction of intensity of contact condensation of steam from the steam–gas mixture by ∼29% in the experiment with nitrogen and by ∼57% in the experiment with helium. The obtained experimental data can be used for numerical simulation of emergency processes in the VVER reactor facility during operation of passive safety systems taking into account the removal of steam–gas mixture from steam generator by supplying the subcooled liquid jet into the volume HA-2 accumulator tanks.

Identification of accident sequences for the DEMO plant

Safety studies are performed in the frame of the conceptual design studies for the European Demonstration Fusion Power Plant (DEMO) to assess the safety and environmental impact of design options. An exhaustive set of reference accident sequences are defined in order to evaluate plant response in the most challenging events and compliance with safety requirements.The Functional Failure Mode and Effect Analysis (FFMEA) has been chosen as analytical tool, as it is a suitable methodology to define possible accident initiators when insufficient design detail is available to allow for more specific evaluation at component level. The main process, safety and protection functions related to the DEMO plant are defined through a functional breakdown structure (FBS). Then, an exhaustive set of high level accident initiators is defined referring to loss of functions, rather than to specific failures of systems and components, overcoming the lack of detailed design information. Nonetheless reference to systems or main components is always highlighted, as much as possible, in order to point out causes and safety consequences.Through the FFMEA a complete list of postulated initiating events (PIEs) is selected as the most representative events in terms of challenging conditions for the plant safety.All the four blanket concepts of the European DEMO reactor have been analysed.

Mines Systems Safety Improvement Using an Integrated Event Tree and Fault Tree Analysis

Mines systems such as ventilation system, strata support system, flame proof safety equipment, are exposed to dynamic operational conditions such as stress, humidity, dust, temperature, etc., and safety improvement of such systems can be done preferably during planning and design stage. However, the existing safety analysis methods do not handle the accident initiation and progression of mine systems explicitly. To bridge this gap, this paper presents an integrated Event Tree (ET) and Fault Tree (FT) approach for safety analysis and improvement of mine systems design. This approach includes ET and FT modeling coupled with redundancy allocation technique. In this method, a concept of top hazard probability is introduced for identifying system failure probability and redundancy is allocated to the system either at component or system level. A case study on mine methane explosion safety with two initiating events is performed. The results demonstrate that the presented method can reveal the accident scenarios and improve the safety of complex mine systems simultaneously.

Recent developments in IFE safety and tritium research and considerations for future nuclear fusion facilities

Over the past five years, the fusion energy group at Lawrence Livermore National Laboratory (LLNL) has made significant progress in the area of safety and tritium research for Inertial Fusion Energy (IFE). Focus has been driven towards the minimization of inventories, accident safety, development of safety guidelines and licensing considerations. Recent technology developments in tritium processing and target fill have had a major impact on reduction of tritium inventories in the facility. A safety advantage of inertial fusion energy using indirect-drive targets is that the structural materials surrounding the fusion reactions can be protected from target emissions by a low-pressure chamber fill gas, therefore eliminating plasma-material erosion as a source of activated dust production. An important inherent safety advantage of IFE when compared to other magnetic fusion energy (MFE) concepts that have been proposed to-date (including ITER), is that loss of plasma control events with the potential to damage the first wall, such as disruptions, are non-conceivable, therefore eliminating a number of potential accident initiators and radioactive in-vessel source term generation.In this paper, we present an overview of the safety assessments performed to-date, comparing results to the US DOE Fusion Safety Standards guidelines and the recent lessons-learnt from ITER safety and licensing activities, and summarize our most recent thoughts on safety and tritium considerations for future nuclear fusion facilities.

ASSESSMENT OF RELIABILITY AND RISK DEGREE FOR ACCIDENT INITIATION AT SLIME STORAGES OF 4th MINING ADMINISTRATION, JSC “BELARUSKALI”

Definition of reliability for dams of slime storage embankment is given on the basis of reliability theory and characteristics of reliability and their analysis are presented in the paper. The paper specifies qualitative indices for earth dams which are subdivided in two groups: applicability factors and structural reliability factors. A short analysis of all possible causes for accident initiation at earth dams has been made and the analysis has permitted to pinpoint eleven main objects for diagnosis for slime storage dams. In order to assess risk degree of accident initiation at JSC “Belaruskali” slime storages all possible causes of emergency cases and their probability of occurrence have been analyzed in the paper. The paper acknowledges the fact that dam malfunction is possible, as a rule, due to violation of operational rules and regulations. Main parameters of slime storage state which are to be controlled regularly in the process of its operation have been noted in the paper. Observation results over slime storages, calculations of dam slope stability for normal operation (a principal calculation case) and operating irregularities in water seals (a special calculation case). As a stability margin factor is close to 1.0 for a special calculation case, an extreme position of depression curve has been determined for all design sections. It has been recommended to carry out a constant control over its position, and in the case when it reaches its peak value it is necessary to undertake appropriate measures in order to reduce its value. Final expert estimations on probability of accident initiation at the investigated slime storage dams of the 4th Mining Administration, JSC “Belaruskali” have been prepared on the basis of the analysis comprising all the required factors. A conclusion has been made about low risk degree of their destruction.

Evaluation of post accident molten material relocation time to the core catcher in a fast breeder reactor

Core Disruptive Accident (CDA) is categorized as a Beyond Design Basis Event in a Fast Breeder Reactor (FBR) because of highly reliable, redundant and diverse plant protection systems. Nonetheless, it is analysed with a view to assess its consequences and take remedial measures to mitigate its consequences. In the present paper, core-melt relocation studies have been carried out to assess the time taken for molten core material to get relocated to core catcher, by melting through various structures beneath the active core. Unprotected Loss of Flow Accident (ULOFA) and Protected Loss of Heat Sink (PLOHS) accident are the bounding events arising from different accident initiators. These are considered with the objective of estimating minimum and maximum time for core-melt relocation to the core catcher for the future Indian FBRs. Numerical heat transfer analysis has been carried out for both the postulated accident scenarios with conservative boundary conditions. In the case of ULOFA where the core-melt is envisaged to reach the grid plate within a few seconds, grid plate melt-through occurs in 300 s for constant molten fuel attack. On the other hand, for a PLOHS scenario, molten core relocation from core mid-plane to core catcher takes place in about 5.5 h with very conservative input conditions.

Key Aspects of the Safety Study of a Water-Cooled Fusion DEMO Reactor

Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. identification of hazardous inventories, identification of stored energies that can mobilize these hazardous inventories and identification of accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant accident (ex-VV LOCA) of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.