Swedish Nuclear Power Plants Research Articles

Numerical assessment for entry condition of severe accident management guidelines in a Swedish nuclear power plant

The entry conditions of severe accident management guidelines (SAMGs) in pressurized water reactors (PWRs) rely on the indication of core exit temperature (CET). Yet, the setpoints for the CET may be different from plant to plant. Most Westinghouse PWR designs adopt the setpoint of CET at 650℃ as the entry condition of the SAMGs, since this setpoint is an effective indicator of core damage in a wide spectrum of accident sequences. Motivated by the interest in the verification & validation of SAMS after the Fukushima accidents, the present study is conducted numerically to verify the effectiveness of the CET setpoint for the transition from emergency operation procedures (EOPs) to SAMGs in a Swedish nuclear power plant. For this purpose, six representative severe accident sequences covering the main contributors to the core damage frequency (CDF) are analyzed using the MELCOR code. Moreover, different CET readings and alternative entry conditions are also investigated. The simulation results show that the average CET = 650 °C is the effective setpoint as the entry condition of SAMGs, i.e., given this setpoint the transition from EOPs to SAMGs will take place slightly before the occurrence of core degradation, which secures the intended mitigation of SAMGs while keeping EOPs active as long as possible. On the other hand, it is too conservative if the maximum CET = 650 °C is used as the setpoint of entry condition of SAMGs, i.e., it will result in an excessive realization of SAMGs over EOPs. The coolant temperature in the primary circuits, the water level in the RPV and the hydrogen concentration in the containment can also be applied as reference indications of core damage states in the accident management.

To trust or not to trust? Structures, practices and discourses of transboundary trust around the Swedish nuclear power plant Barsebäck near Copenhagen

This article examines the role of transboundary trust throughout the evolution of relations regarding the Swedish Barsebäck nuclear power plant, located opposite Copenhagen on the narrow straits separating Denmark and Sweden, which created cross-border tensions until its closure in 1999/2005. The article examines changes in transboundary trust in and between the actors relevant for decision-making regarding the plant, during the three phases of the conflict. It focuses on three aspects, which shaped the development of transboundary trust: first, the role played by pre-existing structures of cross-border institutional, interpersonal and ideological trust, secondly, practices of trust-building that various actors applied to maintain, enhance or rebuild transboundary trust, and thirdly, the growing discourses of mistrust and even distrust in the institutions advocating the use of nuclear power – regardless on which side of the border they were located. The article argues that changes in contemporary views on nuclear risk and the concomitant politicisation of the issue of nuclear power broadened the range of relevant actors, and contributed to changing trust relations. It suggests a mutually reinforcing interaction between interpersonal, institutional and ideological trust, and finds few differences in discourses about institutional trust between transboundary and domestic authorities. Its findings highlight the importance of productive mistrust for nuclear safety, and thus demonstrate that both trust and mistrust are important for living with, but also for limiting (nuclear) risk.

Averting cumulative lifetime attributable risk (LAR) of cancer by decontamination of residential areas affected by a large-scale nuclear power plant fallout: time aspects of radiological benefits for newborns and adults

The averted cumulative lifetime attributable risk (LAR), the residual dose and highest ground deposition of 137Cs complying with a reference dose level of 20 mSv yr−1 to an individual returning after one year to an area contaminated by unfiltered releases of fission products from a nuclear power plant (NPP) were evaluated by applying an existing exposure model designed to compute age- and gender-dependent time-integrated LAR. The model was applied to four types of nuclear fallout scenarios, partly based on data from the Chernobyl and Fukushima releases and from theoretical source terms from Swedish NPPs. For rapid decontamination measures that achieve a 50% relative reduction in external dose rate within 1 year, compliance with the reference level 20 mSv yr−1 can be attained for an initial 137Cs ground deposition of up to 2 MBq m−2 with relaxed food restrictions. This compliance can be attained at even higher ground deposition (up to 4.5 MBq m−2) if using the strict food restrictions employed in Japan after 2011. Considering longer than 1 year return times it was also found that the benefit of implementing decontamination decreases rapidly with time (2–3 years half-time), especially if the fallout has a high initial 134Cs to 137Cs activity ratio and if the ecological half-time of the external dose rate is short (<5 years). Depending on fallout scenario the averted cumulative LAR for newborn girls by decontamination that is achieved after 5 years is only between 6% and 11% of that obtained by evacuation alone during the same time, indicating a rather limited radiological benefit of decontamination if delayed more than a few years. We conclude that decision makers and emergency response planners need to consider that protracted decontamination measures may have limited radiological benefit compared with evacuation in terms of averted future cancer cases, albeit it may have other societal benefits.

Human factors methods for early evaluation of control room systems – guidelines for use in practice

Formative human factors (HF) evaluation is an important activity in control room design. Its purpose is to provide design feedback in the development process, preferably early so as to lower the risk of late and sub-optimal design changes. Early evaluation requires assessment of higher level design decisions due to the gradual specificity of design decisions made as the development process progresses. There is also a need to further develop HF methods that are applicable in practice. The purpose of this article is to seek understanding of, and base guidelines on, the practical use of HF methods for formative assessment of higher level design decisions in control rooms. Evaluations in three development projects at a Swedish nuclear power plant were studied, identifying 18 method guidelines. The guidelines related to: 1) the method’s ability to define, adapt, focus and balance; 2) execution of the evaluation workshops; 3) the communicative purpose of the evaluation activity. Comparing the identified guidelines with literature showed similarities, but also highlighted unique guidelines. The compiled results can be used to further develop methods for early formative evaluation of control rooms in practice.

Radiological environment within an NPP after a severe nuclear accident

The radiological environment following a severe nuclear accident can be visualised on building layouts. The direct radiation in an area (or room) can be visualized on the layout by a colouring scheme depending on the dose rate level (for example orange for high gamma dose rate level and purple for an intermediate gamma dose rate level). Following the Fukushima accident, a need for update of these layouts has been identified at the Swedish nuclear power plant of Forsmark. Shielding calculations for areas where access is desired for severe accident management have been performed. Many different sources of radiation together with different types of shielding material contribute to the dose that would be received by a person entering the area. External radiation from radioactivity within e.g. pipes and components is considered and also external radiation from radioactivity in the air (originating from diffuse leakage of the containment atmosphere). Results are presented as dose rates for relevant dose points together with a method for estimating the dose rate levels for each of the rooms of the reactor building.

Experience feedback from in-depth event investigations: How to find and implement efficient remedial actions

The present research focuses on the processes of identifying remedial actions subsequent to incidents at two Swedish nuclear power plants. Data from 106 in-depth analyses were analysed together with interviews with event investigators. The results and previous research in the domain indicated a need to further develop the process for identifying remedial actions. A method was developed that focuses on process descriptions and identifications of strengths and weaknesses inherent in the process(es) in which an incident occurred. The method uses a participatory approach with actors from the relevant process(es). A case study was conducted which showed promising results. The method is discussed in terms of generalising to a more process-oriented experience feedback than usually is applied.

A belief function theory based approach to combining different representation of uncertainty in prognostics

In this work, we consider two prognostic approaches for the prediction of the Remaining Useful Life (RUL) of degrading equipment. The first approach is based on Gaussian Process Regression (GPR) and provides the probability distribution of the equipment RUL; the second approach adopts a Similarity-Based Regression (SBR) method for the RUL prediction and belief function theory for modeling the uncertainty on the prediction. The performance of the two approaches is comparable and we propose a method for combining their outcomes in an ensemble. The least commitment principle is adopted to transform the RUL probability density function supplied by the GPR method into a belief density function. Then, the Dempster’s rule is used to aggregate the belief assignments provided by the GPR and the SBR approaches. The ensemble method is applied to the problem of predicting the RUL of filters used to clean the sea water entering the condenser of the boiling water reactor (BWR) in a Swedish nuclear power plant. The results by the ensemble method are shown to be more satisfactory than that provided by the individual GPR and SBR approaches from the point of view of the representation of the uncertainty in the RUL prediction.

In the Eye of the Beholder: On Using Photography in Research on Sustainability

This study of perspectives and opinions on a Swedish nuclear power plant and an old building in a Swedish city center employed an ethnographic approach. Such methods are found to be applicable to research situations when it is of crucial importance to establish trustworthy relationships with the informants. The use of ethnographic research methods, especially photography, enabled the researchers in the project to collect a wide range of narratives about different matters regarding the role of the sites and buildings in the past, future prospects of the power plant and the old building and their location. By employing qualitative methods, the project aimed at gaining insight into the meaning-building processes of the actors involved and how they made sense of a place that was turned into a locality of energy production and city planning. This paper elaborates on photography as part of an ethnographic approach and the authors argue that photography helps the researchers to extract the seen and unseen as well as values connected to sustainability in the daily life of local actors.

Oxidation of nickel-base welds 182 and 82 in simulated primary water of pressurised water reactors

To date, nickel-base welds 182 and 82 used in pressurised water reactors have shown a higher susceptibility to stress corrosion cracking (SCC) during laboratory tests than in power plants. However, the increasing number of cracks reported in American, Swedish and Japanese nuclear power plants on Alloys 182/82 enlightens the need for a predictive initiation model of SCC. Initiation of SCC involves several variables interacting together such as material microstructure and composition, primary water chemistry and loading history. Building such a model requires studying each contribution of individual phenomenon, in order to have a better understanding of the mechanisms involved at a local scale in crack initiation. Before SCC initiation, some grain boundaries exposed to primary water present intergranular oxidation reaching several hundreds of nanometres. The oxide penetration kinetics could be modified by several parameters such as local metallurgy, dissolved hydrogen content of the primary water and mechanical history before oxidation. The grain boundaries affected by oxidation during a tensile test demonstrate a lower resistance to failure compared to ‘healthy’ grain boundaries. The main objective of this study is to determine key local parameters controlling grain boundary oxidation in order to calibrate an oxidation model which will be integrated to SCC initiation modelling.

Aspects of the Nuclear Power Plant Control Room System Contributing to Safe Operation

The performance of a nuclear power plant's operational steering entity, the control room system, has a major impact on operational safety. The construction and modernization of nuclear power plants create a need to evaluate control room systems to assess if they fulfil their purpose to support safe operation. The aim of this study was to identify a foundation for evaluation measures, i.e. to find aspects of the control room system that contribute to safe operation from a human factors perspective. A number of professionals in roles influencing the operation and/or design of Swedish nuclear power plants were interviewed: reactor operators, shift supervisors, instructors, human reliability analysis specialists, human factors specialists, and personnel from the Swedish Radiation Safety Authority. A number of aspects contributing to safe operation were identified and categorised in six overall themes: situations, functions, tasks, structural elements, and characteristics. Situations describe states of and/or events in the surrounding environment that the control room system must be able to handle. Functions are the abilities the control room system must have, and tasks are what operators or technical systems in the control room system must be able to perform. Structural elements are the entities that constitute the control room system, and the characteristics of the structural elements establish conditions for the design of artefacts as well as the behaviours and abilities of personnel. These aspects are examples of what is required for controlled performance of the control room system, a prerequisite for safe operation. Together the themes serve as a structure for defining evaluation measures. The themes can be considered variables or constants when deciding on control room system evaluation measures, a choice that must be made for each evaluation. User experience aspects contribute to safe operation, but the operators’ well-being has a value in its own right and should be considered as an additional goal when designing and evaluating control rooms.

The Fukushima Daiichi nuclear accident.

1. Introduction I remember vividly the event in 1986 when the Unit 4 reactor at (1) exploded, since I was working in Russia during the weeks following it. Largely, this was a preventable occurrence, and was caused by a combination of circumstances, but principally through an unscheduled and ill-conceived experiment involving the full withdrawal of the majority of the control rods from the reactor, actually in defiance of standing rules and after deliberately disabling safety systems. In part, the reason for the withdrawal of so many of the control rods was an attempt to compensate for the loss of power caused by a build-up of xenon, which acts as a neutron absorber. Various factors contributed to a loss of cooling water which heightened the already unstable condition of the reactor, due to an increase in the production of steam in the cooling channels (positive void coefficient). By this stage, there was nothing that could be done to avert a calamity, since inherent positive feedback effects rendered the initial rise in power unstoppable, leading to an overwhelming power surge, estimated to be 100 times the nominal power output of the reactor. There is much speculation as to how many deaths might eventually cause, but the initial recorded number was 562 (including 47 liquidators and nine children who died of thyroid cancer). Estimates of the ultimate number range from 4,0002 up to nearly one million (3) fatalities from radiation-induced cancers. is the third really serious nuclear accident to occur in the civilian nuclear industry. There was very little information made available within the USSR, and my Russian colleagues learned most about what had happened from their counterparts in the West. I have mentioned Chernobyl to various of my friends and acquaintances recently, and from this small survey it seems that no-one under the age of about 50 is aware of even the name of the place, let alone what happened there. Apparently, a worker at a Swedish nuclear power plant (NPP) set the alarms off when he went into work on a Monday morning, having been hiking over the weekend. Naturally, this was a surprise since someone working at an NPP might be expected to be contaminated by exposure inside the installation, but in this case, the radioactive plume had contaminated Sweden (and the NPP worker) along with much of Western Europe, which alerted that the event had taken place. Along with Chernobyl, there have been major accidents at Three Mile Island and Windscale (subsequently renamed Sellafield). That said, the nuclear industry has a fairly impeccable safety record, albeit that the long-term storage of its waste remains an unresolved dilemma. To the above list of three, which occurred some decades ago, can now be appended the Fukushima Daiichi nuclear accident (4). In the latter instance, the tsunami following the Tohoku earthquake on 11 March 2011 resulted in failures of equipment and a loss-of-coolant event, with nuclear meltdowns (overheating and damage to the reactor core, with melting of fuel rod components and the potential for escape of radionuclides), and the egress of radioactive isotopes, commencing on 12 March 2011. Fukushima and are the only NPP accidents to measure Level 7 on the International Nuclear Event Scale (5) (described below), and it is estimated that the Fukushima accident has released (6) 10-30% of the amount of radiation resulting from that at Chernobyl, although the emissions continue, and it is not clear how and when they may be stemmed entirely. The Fukushima NPP (7) had six separate boiling water reactors which were made by General Electric (GE) and maintained by the Tokyo Electric Power Company (TEPCO). Reactor 4 had been de-fuelled when the incident took place, and reactors 5 and 6 were in cold shutdown with the intention of a planned maintenance effort. When the earthquake hit, reactors 1-3 were automatically shut down by the insertion of control rods, i. …

Reactive transport modelling of organic complexing agents in cement stabilized low and intermediate level waste

The Swedish final repository for short-lived radioactive waste (SFR 1) is located at Forsmark in Sweden. It holds low and intermediate-level operational waste from the Swedish nuclear power plants, as well as industrial, research-related, and medical waste. A variety of low molecular weight organic compounds are present in the waste or in its matrix. Such compounds can also be formed by chemical degradation of organic macromolecules. These organics can ligate to metal atoms forming stable complexes and also adsorb to the surface of cement, thereby influencing the net release of radionuclides from the repository. This motivates the study of the concentration distribution of complexing agents in the repository as a function of time. The following paper reports the results of mass transport modelling, describing the transport of complexing agents through the cementitous matrix in the rock vault for intermediate-level waste in the SFR 1 repository. Nitrilotriacetate (NTA) and isosaccharinate (ISA) have been investigated, where the former is considered to be non-sorbing and non-reacting, while the latter is produced from cellulose degradation and adsorbs strongly to cement. The 3D model considers advection, diffusion, and sorption of solvated species in cement pore water over a time period of 20,000years. The model accounts for the spatial distribution of the flow field in the repository structure and also considers changing groundwater flow during the investigated time period. It is found that 99% of the NTA is removed after approximately 4000years, while 90% of the ISA is retained in the rock vault after 20,000years. The maximum pore water concentration of ISA is found to be 8.6mol/m3 after approximately 2300years, based on the degradation of the deposited amounts of cellulose. Over the investigated time scale, the ligands retained in the repository can redistribute across several waste compartments where the organic compounds were not originally deposited. In the case of ISA this effect is dampened due the influence of sorption.

Effects of staff bonus systems on safety behaviors

Bonus systems are a common means in trying to motivate employees to perform well. However, there is still disagreement regarding the effects of bonus systems. Some theories even suggest that such systems could cause an increase in risk-behavior. This makes further research regarding bonus systems warranted, especially when applied in high-risk organizations. This study aims to explore potential effects on safety-related behavior associated with bonus systems currently used at Swedish nuclear power plants. Fifteen semi-structured interviews with employees were performed based on an eclectically composed framework from motivational and organizational theories. Results do not indicate any negative effects on safety-related behaviors, but rather that safety behaviors may be promoted insofar as bonus rewards are linked to performance goals concerning safety. Differences in bonus system design appeared to affect behavioral outcomes. The comparative and qualitative approach of this study contributes valuable information by highlighting the types of factors that may serve to stimulate greater incentive for employees to engage in safe behavior.

Galvanic corrosion properties of steel in water saturated concrete

Aluminum-based sacrificial anodes were installed to reinforced concrete to stop ongoing corrosion in cooling water tunnels in a Swedish nuclear power plant. The steel rebars were also unintentionally connected to stainless steel water pumps. Therefore, the consumption rate of the sacrificial anodes was higher than predicted. An experimental and a field study were performed to assess if the steel rebar suffer from galvanic corrosion and if the stainless steel pumps are responsible alone for the high consumption rate. It was found from the experimental study that there is an increased risk of galvanic corrosion for steel rebar when the corrosion potential is raised to -200mV (SCE) for samples with 1% Cl- by mass of cement and -500mV (SCE) for samples with 2% Cl- by mass of cement. The experimental results were compared with the corrosion potential measured in the cooling water tunnel where sacrificial anodes were in use and not mounted at all. The cooling water tunnel without sacrificial anodes had generally more anodic corrosion potentials compared to the tunnel with anodes. The tunnel with anodes had also more anodic potentials closer to the stainless steel pumps than further away which means that the rebar is affected by the stainless steel pumps. However, the measured corrosion potentials in the tunnels were not as anodic as the potentials needed for high galvanic current measured in the experimental study. (Less)

Professional subcultures in nuclear power plants

Using a safety climate survey as the point of departure, the present study explores some aspects of plant cultures vs. professional subcultures in three Swedish nuclear power plants (named A, B and C). The ratings on the safety climate survey by workers on power plant A were subjected to an exploratory factor analysis. A six-factor solution explained a total of 56.0% of the variance in the items included. The six factors were considered to measure Safety management, Change management and experience feedback, Immediate working group, Knowledge and participation, Occupational safety, and Resources. The six factor model was tested by running a confirmatory factor analysis on the ratings by workers on power plant B and C, respectively. The model fit for both plants was acceptable and supported the six factor structure. For each of the six factors, a 3×3 ANOVA was conducted on the ratings, with the three largest departments (Operation, Maintenance, Engineering support) and power plants (A, B, C) as the between-subjects factors. Differences between power plants as well as differences between departments were found for several factors. Overall, the differences between departments were larger than those between power plants. The results are discussed in terms of challenges for creating safety climate in organizations that harbor several professional subcultures.

Less is more? Results from a case study on improving the safety review process at a nuclear power plant

This paper focuses on technical safety reviews of plant modifications at a Swedish nuclearpower plant. The primary aims of the presented study include identifying the main areas forimprovement of t ...

Identification of the chemical inventory of different paint types applied in nuclear facilities

The floors, concrete walls and many of the metal surfaces in nuclear power plant containments are coated with zinc primers or paint films to preserve the metal surfaces and simplify decontamination in the containment after the occurrence of a severe nuclear incident or accident. A chemical examination of paint films from different nuclear installations out of operation, as well as current operating ones, reveals that different types of paints are used whose composition can vary significantly. Results obtained for one type of paint at a certain nuclear site are in most cases unlikely to be comparable with sites painted with another type of paint. During normal operation and particularly during nuclear accidents, the paints will degrade under the high temperature, steam and irradiation influence. As paint and its degradation products can act as sources and depots for volatile iodine compounds, the type and aging conditions of the paint films will have a significant impact on the source term of the volatile fission product iodine. Thus, great care should be taken when extrapolating any results obtained for the interaction of radioactive iodine with one paint product to a different paint product. The main focus of the study is a comparison of the chemical profile of paint films applied in Swedish nuclear power plants. Teknopox Aqua V A, an epoxy paint recently used at Ringhals 2, and an emulsion paint used in the scrubber buildings of Ringhals 1-4 are compared with a paint film from Barseback nuclear power plant unit 1 that had been aged under real reactor conditions for 20 years. In addition, two paint films, an emulsion and a gloss paint, used in an international nuclear fuel reprocessing facility, are compared with the paints from the Swedish nuclear power plants.

Challenges in performing technical safety reviews of modifications – A case study

The aim of the present study, is to identify strengths and weaknesses of the technical safety review process at a Swedish Nuclear Power Plant (NPP). In this context, the function of safety reviews are understood as expert judgements on proposals for design modifications1“Design modifications” are here understood as alterations of an existing design.1 and redesign of technical systems (i.e. commercial nuclear reactors), supported by formalised safety review processes. The chosen methodology is using two complementary methods: interviews of personnel performing safety reviews, and analysis of safety review reports from 2005 to 2009.The study shows that personal integrity is a trademark of the review staff and there are sufficient support systems to ensure high quality. The partition between primary and independent review is positive, having different focus and staff with different skills and perspectives making the reviews, which implies supplementary roles. The process contributes to “getting the right things done the right way”. The study also shows that though efficient communication, feedback, processes for continuous improvement, and “learning organizations” are well known success factors in academia, it is not that simple to implement and accomplish in real life.It is argued that future applications of safety review processes should focus more on communicating and clarifying the process and its adherent requirements, and improve the feedback system within the process.

Simulation based validation of the detection capacity of an ultrasonic inspection procedure

With aging engines and structures, reliable maintenance will be a very important challenge in the future. This and other safety aspects have encouraged the development of non-destructive testing (NDT) technologies to detect in-service defects at an early stage. In order to quantify the inspection reliability the methodology of probability of detection (POD) was developed by the aeronautical industry in the early 1980s. This statistical tool reduces the number of artificially produced artefacts that needs to be introduced into the test blocks in order to get statistically valid information of the detection capacity, even though its legitimacy is limited to very restricted conditions.Even though this procedure reduces the number of defect it includes a large number of inspections and personnel and these campaigns thus tend to be time consuming and very expensive. This paper address the development of a procedure for generating POD based on synthetic data using NDT simulation software. The intention is to have an optimized experimental phase combined with much more efficiently retrieved simulated data. This has been achieved by fitting a multi-parameter prediction model to ultrasonic simulation software (simSUNDT) in an orthogonal design of experiments.The probabilities of detection as function of defect size (POD curves) at different defect depths were generated by Monte Carlo simulation introducing variations in the control factors with a physical interpretation in the emulator.The validation of this developed methodology has been based on a qualified ultrasonic procedure, UT-01. It specifies manual ultrasonic inspection of piping components within Swedish nuclear power plants and qualification of personnel according to this procedure has been in place since 1996. Experiences made by these qualifications were comprised as POD curves for both fatigue cracks and stress corrosion cracks in 2005. These POD curves are in this paper compared with corresponding generated with the above described simulation based methodology and the agreement found was remarkable well.

Analyses of loss-of-coolant accidents involving release of insulation material under consideration of the measures taken for PWR from expert's point of view

Abstract The incident, which happened in the Swedish nuclear power plant Barsebäck 2 on 28th of July, 1992 brought about a re-evaluation of the design requirements for the control of loss-of-coolant accidents (LOCA) involving the release of insulation material that were valid in Germany up to that time. The current paper provides an overview on the aspects that need to be taken into consideration for analyses as well as a description of the important boundary conditions. In giving details of the analyses, the actual state is also described. It is also illustrated that sufficient examinations and tests have been performed for the German PWR plants and corresponding measures have been taken providing a sufficient basis for a final positive evaluation of the issue “LOCA involving release of insulation material”.