List all NKS-R reports

Under the NKS-R programme, activities are grouped under the following research areas:

Thermal hydraulics
Severe accidents
Reactor physics
Risk analysis & probabilistic methods
Organisational issues and safety culture
Decommissioning, including decommissioning waste
Plant life management and extension

	Data4decom
	Source Term and Timing Uncertainty in Severe accidents NKS-STATUS Phase 3 report
	Barsebäck as a Research and Development Platform, Extraction and Analysis of Service-aged and Irradiated Reactor Pressure Vessel Material
	Effect of Thermal Ageing and Microstructure on Fracture Mechanical Behaviour of Ni-based Alloy Dissimilar Metal Welds
	Internal nuclear safety oversight as part of organizational defence-in-depth – Lessons learned for the Nordic nuclear industry - Intermediate report from the NKS-R INSOLE activity
	Source Term and Timing Uncertainty in Severe accidents NKS-STATUS Phase 2 report
	Position-sensitive detectors for nuclear fuel imaging (POSEIDON)
	Organic Telluride Formation during a Severe Nuclear Accident (ORTEF)
	Barsebäck as a Research and Development Platform, Extraction and Analysis of Service-aged and Irradiated Reactor Pressure Vessel Material
	Effect of Thermal Ageing and Microstructure on Fracture Mechanical Behaviour of Ni-based Alloy Dissimilar Metal Welds
	Thermal Hydraulic Phenomena of the Suppression Pool
	Micromechanical modeling of single crystal and polycrystalline UO2 at elevated temperatures
	Source Term And Timing Uncertainty in Severe Accidents
	Corrosion of copper in sulphide containing environment: the role and properties of sulphide films – Annual report 2021
	Barsebäck as a Research and Development Platform, Extraction and Analysis of Service-aged and Irradiated Reactor Pressure Vessel Material
	Thermal Hydraulic Phenomena of the Suppression Pool - Summary of 2020 activities
	Corrosion of copper in sulphide containing environment: the role and properties of sulphide films – Annual report 2020
	WPS-MAF, Report on design and results from experimental test program for material characterization
	Towards high-fidelity fuel pellet fracture modelling in current and new fuel designs
	Prolonged Available Time and Safe States
	Barsebäck as a Research and Development Platform, Extraction and Analysis of Service-aged and Irradiated Reactor Pressure Vessel Material
	Scenarios and Phenomena Affecting Risk of Containment Failure and Release Characteristics
	Thermal Hydraulic Phenomena of the Suppression Pool
	Main Outcomes of the TETRA Project on Tellurium Chemistry in a Severe Accident
	Corrosion of copper in sulphide containing environment: the role and properties of sulphide films - Annual report 2019
	Prolonged Available Time and Safe States
	Barsebäck as a Research and Development Platform, Extraction and Analysis of Service-aged and Irradiated Reactor Pressure Vessel Material; Summary of 2019 activities
	Scenarios and Phenomena Affecting Risk of Containment Failure and Release Characteristics
	WRANC, Warm Pre-Stressing – Validation of the relevance of the main mechanisms behind Warm Pre-Stressing in assessment of nuclear components
	Determination of Fire Barriers reliability for fire risk assessment of Nuclear Power Plants. (FIREBAN) – Final Report
	Synthetic ground motions to support the Fennoscandian GMPEs
	Site risk analysis for nuclear installations
	Barsebäck as a Research and Development Platform, Extraction and Analysis of Service-aged and Irradiated Reactor Pressure Vessel Material
	Challenges and opportunities for improving Nordic nuclear decommissioning
	Development of a hybrid neutron transport solver in 2 energy groups – final report
	Phenomena Threatening Containment Integrity in Deterministic and Probabilistic analyses and integration of PRA levels 1 and 2
	Scenarios and Phenomena Affecting Risk of Containment Failure and Release Characteristics
	COPSAR-NKS Project; Summary report 2017
	Phenomena Threatening Containment Integrity and Evaluating Source Term Characteristics
	Safety Culture Assurance and Improvement Methods in Complex Projects – Final Report from the NKS-R SC_AIM
	Challenges and opportunities for improving Nordic nuclear decommissioning
	Determination of Fire Barriers reliability for fire risk assessment of Nuclear Power Plants (FIREBAN) – Report Year 1
	ATR-2 Part B: Ruthenium transport in an RCS with airborne CsI
	Scenarios and Phenomena Affecting Risk of Containment Failure and Release Characteristics
	Modelling as a Tool to Augment Ground Motion Data in Regions of Diffuse Seismicity – Final report
	Modelling of a Large Water Pool during Operation of Blowdown Pipes, Spargers, and Nozzles
	Simulation of PPOOLEX stratification and mixing experiment SPA-T1
	Data and visualization solutions for HYBRID core simulation method
	Development of a hybrid neutron transport solver in 2 energy groups
	Addressing off-site consequence criteria using Level 3 PSA
	Barsebäck as Research and Development Platform, Extraction and Analysis of Reactor Pressure Vessel Material
	Preliminary Spray Tests in PPOOLEX
	Mixing Tests with an RHR Nozzle in PPOOLEX
	Sparger Tests in PPOOLEX on the Behaviour of Thermocline
	Safety Culture Assurance and Improvement Methods in Complex Projects – Intermediate Report from the NKS-R SC_AIM
	Addressing off-site consequence criteria using Level 3 PSA - Phase 3 Status Report from the NKS-R L3PSA
	Ex-Vessel Steam Explosion Analysis with MC3D
	A summary of studies on debris bed coolability and multi-dimensional flooding
	ATR-2 Part A: Ruthenium chemistry and transport in a RCS due to air radiolysis products
	Experimental and Analytical Investigations of Debris Bed Formation, Spreading, Coolability, and Steam Explosion in Nordic BWRs
	Thermal Stratification and Mixing in a Large Pool Induced by Operation of Spargers, Nozzles, and Blowdown Pipes
	Pre-calculation of a PPOOLEX spray experiment
	Single Spray Nozzle Tests
	CFD simulations of preliminary single spray experiments
	Additional Sparger Tests in PPOOLEX with Reduced Number of Injection Holes
	Modelling as a tool to augment ground motion data in regions of diffuse seismicity - Progress 2015
	Modelling of DIgital I&C, MODIG — Interim report 2015
	Safety Demonstration Planning for Digital I&C Projects – Challenges, Future Directions and Improving Guidance
	Learning from Successes in Nuclear Power Plant Operation - Intermediate Report from the NKS-R LESUN
	MC3D simulations of ex-vessel steam explosions in IDPSA framework
	Treatment of ex-vessel debris coolability in IDPSA context
	Modelling of the Effects of Steam Injection through Spargers on Pool Thermal Stratification and Mixing
	Analysis of Debris Bed Formation, Spreading, Coolability, and Steam Explosion in Nordic BWRs
	Deterministic-Probabilistic Safety Analysis Methodology for Analysis of Core Degradation, Ex-vessel Steam Explosion and Debris Coolability
	Analysing Steam Explosions with MC3D
	Coolability analyses of heap-shaped debris bed
	Software reliability analysis for PSA: failure mode and data analysis
	CFD and FEM modeling of blowdown of gas into pressure suppression pool
	Addressing off-site consequence criteria using Level 3 PSA
	PPOOLEX Experiments with a Sparger
	PIV Measurements of DCC-06 and DCC-07 PPOOLEX Experiments
	Measuring Procedure Competence. Final Report from the NKS-R(14)112/13
	Guidelines for reliability analysis of digital systems in PSA context — Final report
	Nordic Nuclear Forum for Generation IV Reactors 2014
	Human performance tools in nuclear power plant maintenance activities - Final report of HUMAX project
	Prediction and validation of pool fire development in enclosures by means of CFD Models for risk assessment of nuclear power plants (Poolfire) – Final Report
	Human-Performance Tools in Maintenance Work - A Case Study in a Nordic Nuclear Power Plant
	Sensitivity Study of Steam Explosion Energetics in Reference Nordic BWR using MC3D Code
	Study and comparison of data on Coolability of Particulate Beds Packed with Irregular Multi-size Particles
	Analysis of Debris Bed Formation, Spreading, Coolability, and Steam Explosion in Nordic BWRs
	Effective Models for Simulation of Thermal Stratification and Mixing Induced by Steam Injection into a Large Pool of Water
	Feasibility Study for Connection Between IDPSA and conventional PSA Approach to Analysis of Nordic type BWR’s
	A study on the coolability of debris bed geometry variations using 2D and 3D models
	COOLOCE-12 debris bed coolability experiment: Cone on a cylindrical base
	COOLOCE debris bed coolability experiments with an agglomerate
	Chugging Test with DN100 Blowdown Pipe in the PPOOLEX Facility
	PPOOLEX Mixing Experiments
	CFD and FEM simulations of pressure suppression pool behaviour
	Evaluation of Existing Applications and Guidance on Methods for HRA – EXAM-HRA - Phase 3a Summary Report
	Software reliability analysis for PSA
	Addressing off-site consequence criteria using Level 3 PSA — Phase 1 Status Report
	Guidelines for reliability analysis of digital systems in PSA context — Phase 4 Status Report
	Improving design processes in the nuclear domain. Insights on organisational challenges from safety culture and resilience engineering perspectives
	The expected and experienced benefits of Human performance tools in nuclear power plant maintenance activities - Intermediate report of HUMAX project
	NKS-R Decommissioning Seminar 2013
	Simulations of Ex-vessel Fuel Coolant Interactions in a Nordic BWR using MC3D Code
	Study on Effective Particle Diameters and Coolability of Particulate Beds Packed with Irregular Multi-size Particles
	Investigation of debris bed formation, spreading and coolability
	COOLOCE debris bed experiments and simulations investigating the coolability of cylindrical beds with different materials and flow modes
	Adsorption and revaporisation studies of thin iodine oxide and CsI aerosol deposits from containment surface materials in LWRs
	Validation of Effective Momentum and Heat Flux Models for Stratification and Mixing in a Water Pool
	Modelling of pressure loads in a pressure suppression pool
	Prediction and validation of pool fire development in enclosures by means of CFD Models for risk assessment of nuclear power plants (Poolfire) - Report Year 2
	PPOOLEX EXPERIMENTS ON THE DYNAMICS OF FREE WATER SURFACE IN THE BLOWDOWN PIPE
	PIV MEASUREMENTS AT THE BLOWDOWN PIPE OUTLET
	Final report of MoReMO 2011-2012. Modelling Resilience for Maintenance and Outage
	Safety culture in design
	Guidelines for reliability analysis of digital systems in PSA context — Phase 3 Status Report
	Adsorption and revaporisation studies on iodine oxide aerosols deposited on containment surface materials in LWR
	Nordic Forum for Generation IV Reactors, Status and activities in 2012
	Effective Momentum and Heat Flux Models for Simulation of Stratification and Mixing in a Large Pool of Water
	Modeling of interaction of multiple vent pipes in a pressure suppression pool
	Organizational factors in design and implementation of technological and organizational solutions in the nuclear industry.
	Intermediate report of MoReMO. Modelling Resilience for Maintenance and Outage.
	Guidelines for reliability analysis of digital systems in PSA context — Phase 2 status report
	Prediction and validation of pool fire development in enclosures by means of CFD (Poolfire) Report – Year 1
	NOMAGE4 activities 2011, Part II, Supercritical water loop
	NOMAGE4 activities, Part I, Nordic Nuclear Materials Forum for Generation IV Reactors: Status and activites in 2011
	Numerical modelling of pressure suppression pools with CFD and FEM codes
	Development and Validation of Effective Models for Simulation of Stratification and Mixing Phenomena in a Pool of Water
	Wire System Ageing Assessment and Condition Monitoring (WASCO)
	In-vessel Coolability and Steam Explosion in Nordic BWRs
	Experimental study on iodine chemistry (EXSI) - Containment experiments with methyl iodide
	Multiple Blowdown Pipe Experiments with the PPOOLEX Facility
	PPOOLEX Experiments on Stratification and Mixing in the Wet Well Pool
	Nuclear Safety Culture in Finland and Sweden - Developments and Challenges
	Automation strategies in five domains - A comparison of levels of automation, function allocation and visualisation of automatic functions
	CFD and FEM modeling of PPOOLEX experiments
	PPOOLEX Experiments on Dynamic Loading with Pressure Feedback
	PPOOLEX Experiments with Two Parallel Blowdown Pipes
	Comparison of VNEM to Measured Data from Ringhals Unit 3 (Phase 3)
	Guidelines for reliability analysis of digital systems in PSA context - Phase 1 Status Report
	Human Reliability Guidance - How to Increase the Synergies between Human Reliability, Human Factors, and System Design & Engineering. Phase 2: The American Point of View - Insights of How the US Nuclear Industry Works With Human Reliability Analysis
	Human Reliability Guidance - How to Increase the Synergies between Human Reliability, Human Factors, and System Design & Engineering. Phase 1: The Nordic Point of View - A User Needs Analysis
	Modeling of Condensation, Stratification, and Mixing Phenomena in a Pool of Water
	Guidance to Risk-Informed Evaluation of Technical Specifications using PSA
	Experimental study on iodine chemistry (EXSI) -Containment experiments with methyl iodide
	In-vessel Coolability and Steam Explosion in Nordic BWRs
	Nordic Nuclear Materials Forum for Generation IV Reactors
	Safety Culture in the Finnish and Swedish Nuclear Industries - History and Present
	Comparison of VNEM to Measured Data from Ringhals Unit 3 IACIP, NKS-R-2008-61
	Modelling of blowdown of steam in the pressurized PPOOLEX facility
	Studies on the effect of flaw detection probability assumptions on risk reduction at inspection
	Experimental study on iodine chemistry (EXSI) - Containment experiments with elemental iodine
	Improving Accuracy of the Calculation of In-core Power Distributions for Light Water Reactors
	Stratification issues in the primary system. Review of available validation experiments and State-of the-Art in modelling capabilities (StratRev)
	PPOOLEX Experiments on Wall Condensation
	PPOOLEX Experiments with a Modified Blowdown Pipe Outlet
	PPOOLEX Experiments on Thermal Stratification and Mixing
	Wire System Ageing Assessment and Condition Monitoring (WASCO)
	GOTHIC code simulation of thermal stratification in POOLEX facility
	Probabilistic Safety Goals. Phase 3 - Status Report
	Extremes temperatures and enthalpy in Finland and Sweden in a changing climate
	Levels of Automation and User Control - Evaluation of a Turbine Automation Interface
	MORE: Management of Requirements in NPP Modernisation Projects, Final Report
	Probabilistic Safety Goals. Phase 2 - Status Report
	NPSAG/NKS: Interpretation and Evaluation of the Technical Specification Criteria
	LINGUISTIC ANALYSIS OF NPP INSTRUCTIONS
	CFD Simulation of Air Discharge Tests in the PPOOLEX Facility
	Steam Line Rupture Experiments with the PPOOLEX Test Facility
	Characterizing Experiments of the PPOOLEX Test Facility
	Interactions of RuO4(g) with different surfaces in nuclear reactor containments
	Cost Calculations for Decommissioning and Dismantling of Nuclear Research Facilities
	Conducting organizational safety reviews ? requirements, methods and experience
	MORE: Management of Requirements in NPP Modernisation Projects, Project Report 2007
	RADDA - Comparison of results of three ATWS/ATWC scenarios simulated with the help of POLCA-T and S3K/RELAP5
	Analysis of flow fields, temperatures and ruthenium transport in the test facility
	Ex-Vessel Corium Coolability and Steam Explosion Energetics in Nordic Light Water Reactors
	Wire System Aging Assessment and Condition Monitoring (WASCO)
	Ex-Vessel Coolability and Energetics of Steam Explosions in Nordic Light Water Reactors
	Organisational reviews - requirements, methods and experience. Progress report 2006
	Analysis of Loads and Fluid-Structure Interactions in a Condensation Pool
	Probabilistic Safety Goals. Phase 1 ? Status and Experiences in Sweden and Finland
	Automation Inflicted Differences on Operator Performance in Nuclear Power Plant Control Rooms
	Experiments on the behaviour of ruthenium in air ingress accidents
	A Procedure to Generate Input Data of Cyclic Softening and Hardening for FEM Analysis from Constant Strain Amplitude Fatigue Tests in LCF Regime
	Management of Requirements in NPP Modernisation Projects. Project Report 2006
	Combined Effects Experiments with the Condensation Pool Test Facility
	Cost Calculations for Decommissioning and Dismantling of Nuclear Research Facilities, Phase 1
	MORE. Management of Requirements in NPP Modernisation Projects. Project Report 2005
	NKS-R ExCoolSe Mid-term Report. KTH Severe Accidents Research Relevant to the NKS-ExCoolSe
	Nordic perspectives on safety management in high reliability organizations: Theory and applications
	Wire System Aging Assessment and Condition Monitoring (WASCO)
	Fluid-Structure Interaction Analysis of a Water Pool under Loading Caused by Steam Injection
	Simulation and Analysis of Data for Enhancing Low Cycle Fatigue Test Procedures
	Management of change in the nuclear industry - Evidence from maintenance reorganizations
	Experiments on the behaviour of ruthenium in air ingress accidents - Progress report
	Thermal stratification experiments with the condensation pool test rig
	Workshop on decommissioning. (Seminarium om avveckling, Risø, 13-15 september 2005)
	Traceability and Communication of Requirements in Digital I&C Systems Development. Final Report.
	Modeling Goals and Functions of Control and Safety Systems - theoretical foundations and extensions of MFM
	Countermeasures and Barriers
	Ex-Vessel Coolability and Energetics of Steam Explosions in Nordic Light Water Reactors - EXCOOLSE Project Report 2004
	Condensation Pool Experiments with Steam Using DN200 Blowdown Pipe
	Maintenance culture and management of change. - Intermediate report 2004.
	Final Report of the 'Nordic Thermal-Hydraulic and Safety Network (NOTNET)' - Project.
	A summary of the Nordic-group conference on safety management, Lund, Sweden, October 28-29, 2004.
	Nordic Nuclear Safety Research (NKS) Seminar on Experience
	Fluid-Structure Interaction Analysis of a Water Pool under Loading Caused by a Condensation-Induced Water Hammer.
	Traceability and Communication of Requirements in Digital I&C Systems Development - Project Report 2004.
	Knowledge Management in Nordic NPPs. Summary report of the findings from the workshop.
	Nuclear Seminar on Nuclear Automation, 5-7 April 2004 in Oskarshamn, Sweden
	Ruthenium Behaviour in Severe Nuclear Accident Conditions - Final Report.
	Pre-Project on Development and Validation of Melt Behavior in Severe Accidents
	Contextual assessment of maintenance culture at Olkiluoto and Forsmark.
	Risk Informed Decision Making - a Pre-Study.
	CCF Model Comparison.
	Preliminary condensation pool experiments with steam using DN80 and DN100 blowdown pipes.
	Numerical analyses of a water pool under loadings caused by a condensation induced water hammer.
	On safety management and nuclear safety.
	Ruthenium behaviour in severe nuclear accident conditions - progress report.
	Traceability and Communication of Requirements in Digital I&C Systems Development. Project Report 2003.
	VALDOR 2003. Values in Decisions On Risk. Proceedings.
	3D Analysis Methods - Study and Seminar.
	Safety Management: A Frame of Reference for Studies of Nuclear Power. Safety Management and Case Studies from Non-Nuclear Contexts
	Barriers, Control and Management. Report from the pilot phase.
	The Possibility and the Effects of a Steam Explosion in the BWR Lower Head on Recriticality of a BWR Core
	Recriticality calculation with GENFLO code for the BWR core after steam explosion in the lower head
	Systematic Analysis of Dependent Human Errors From the Maintenance History at Finnish NPPs - A Status Report
	The Use of Condition Monitoring Information for Maintenance Planning and Decision-Making
	Comparative Study of Approaches to Estimate Pipe Break Frequencies
	Nuclear Safety in Perspective. Final Report of the Nordic Nuclear Safety Research Project SOS-1
	Structural integrity of a reinforced concrete structure and a pipe outlet under hydrogen detonation conditions
	Advances in operational safety and severe accident research. Final report of the Nordic Nuclear Safety Research Project SOS-2
	A risk informed safety classification for a Nordic NPP
	Severe Accident Research and Management in Nordic Countries - A Status Report
	A Study on Hydrogen Deflagration for Selected Severe Accident Sequences in Ringhals 3
	Formation and behaviour of organic iodine
	An analysis of errors of commission in the Nordic nuclear power plants based on plant operating experience
	Safety and Radiation Protection in Waste Management. Final Report of the Nordic Nuclear Safety Research Project SOS-3
	Radioactivity in commercially available metals
	Human factors in maintenance: Development and research in Swedish nuclear power plants
	Identification and communication of uncertainties of phenomenological models in PSA
	Decision criteria in PSA applications
	Syn på kvalitetssäkring vid finländska och svenska kärnkraftverk samt vid Haldenreaktorn
	Gränsöverskridande kommunikation - Problem och lärdomar i kommunikationen över expertområden
	NKS/SOS-1 Seminar on Safety analysis. Report from a seminar held on 22-23 March 2000. Risø National Laboratory, Roskilde, DK
	NKS/SOS-1 Seminarium, 2000-09-25 - Riskkommunikation i Oskarshamn
	Experiences of Storage of Radioactive Waste Packages in the Nordic Countries
	Experience from the comparison of two PSA-studies
	Organisationskulturen inom den finländska Strålsäkerhetscentralens (STUK) avdelning för övervakning av kärnkraftverk. Organisational culture at the Radiation and Nuclear Safety Authority of Finland's department of Nuclear Reactor Regulation.
	Behandling av säkerhetsrelaterade frågor i kärnkraftverkens ledningsgrupper, (Handling of safety issues in the senior management group at the nuclear power plants)
	NKS/SOS-1 Seminarium om Kvalitetssäkring. Rapport från ett seminarium i Ringhals den 16-17 januari 2001. (NKS/SOS-1 Seminar on Quality Assurance in Nordic nuclear reactor plants. Ringhals, January 16-17, 2001)
	Three-dimensional Simulation of Hydrogen Detonations in the Olkiluoto BWR Reactor Building
	Gas Phase Chemistry and Removal of CH3I during a Severe Accident
	Safety- and risk analysis activities in other areas than the nuclear industry
	Current practice for clearance in the Nordic Countries
	Reinforced concrete wall under hydrogen detonation
	Syn på säkerhetskultur vid svenska och finska kärnkraftverk
	On Detonation Dynamics in Hydrogen-Air-Steam Mixtures: Theory and Application to Olkiluoto Reactor Building
	SOS-1 Seminarium om säkerhetskultur. Rapport från ett seminarium i Olkiluoto d. 26-27 oktober 1999
	Miljökonsekvensbeskrivningar vid slut-förvaring av radioaktivt avfall. Temamöte, Roskilde, Danmark, 30-31 augusti 1999
	Final Summary Report of the Nordic Nuclear Safety Research Program 1994-1997
	Methods to Prevent the source Term of Methyl Iodide During a Core Melt Accident
	Reliability Centered Maintenance miniseminar
	Proceedings of the NKS/SOS-2 Seminar on Risk Informed Principles. Bergendal 13.4.-14.4 1999
	Description of the Magnox Type of Gas Cooled Reactor (MAGNOX)
	Säkerhetsindikatorer inom kärnkraftindustrin; definitioner, användning och erfarenheter Rapport från ett seminarium på VTT den 17-18 mars 1999
	Miljökonsekvensbeskrivningar vid slutförvaring av radioaktivt avfall. Temamöte, Gardermoen, Norge, 24-25 november 1998
	Systematisk underhållsanalys med beslutsstöd. Metodik och verktyg för effektivisering av underhållsprogram
	A Procedure for Estimation of Pipe Break Probabilities due to IGSCC
	REAKTORSÄKERHET; En beskrivning och värdering av säkerhetsarbetet i Norden (see also NKS-76)
	Modernisation for Maintaining and Improving Safety at Nordic Nuclear Power Plants
	Integrated Sequence Analysis
	Source Term Calculations - Ringhals 2 PWR
	Parametric Studies on Containment Thermal Hydraulic Loads during high pressure Melt Ejection in a BWR
	Strategies for Reactor Safety: Preventing Loss of Coolant Accidents
	Safety in the Final Disposal of Radioactive Waste
	Safety against Releases in Severe Accidents
	Strategies for Reactor Safety
	Characterisation of long-lived low and intermediate-level radioactive wastes in the Nordic countries
	On core debris behaviour in the pressure vessel lower head of Nordic boiling water reactors
	On recriticality during reflooding of a degrading boiling water reactor core
	Performance analysis for waste repositories in the Nordic countries - Report for Project AFA-1.2
	CAMS Achievements in 1996
	Seminarium om Granskning för säkerhet och kvalitet Strategi och praxis. den 16-17 januari 1997 på VTT i Esbo, Finland
	Studies on melt-water-structure interaction during severe accidents
	En jämförelse mellan myndighetsarbetet inom kärnsäkerheten i Finland och Sverige
	A cognitive task analysis of the SGTR scenario
	CAMS achievements in 1995
	Description of the Prototype Fast Reactor at Dounreay
	Avfallskaraktärisering för långlivat låg- och medelaktivt avfall i Norden - Delrapportering för projekt AFA-1.1
	Coolability of degraded core under reflooding conditions in Nordic boiling water reactors
	Survey of methods for integrated sequence analysis with emphasis on man-machine interaction
	Summary of approaches to dynamic analysis of man-machine interaction
	The CAMS prototype
	Nordic studies in reactor safety
	Safety Evaluation by Living Probabilistic Safety Asssessment and Safety Indicators
	Decommissioning of a Uranium Reprocessing Pilot Plant
	Design and safety features of nuclear reactors neighbouring the Nordic countries
	The Nordic programme on waste and decommisioning (KAN) 1990-93
	Guidance on Clearance from Regulatory Control of Radioactive Materials
	Cleanup of Large Radioactive-contaminated Areas and Disposal of Generated Waste
	Conservation and Retrieval of Information - Elements of strategy to inform future societies about nuclear waste repositories
	Disposal of Radioactive Wastes from the Cleanup of Large Areas Contaminated in Nuclear Accidents - A literature survey
	Aspects of nuclear waste management after a 4-year Nordic programme
	The influence of chemistry on core melt accidents
	Optimization of technical specifications by use of probabilistic methods - A Nordic perspective
	Natural radiation, nuclear wastes and chemical pollutants
	Dependencies, human interactions and uncertainties in probabilistic safety assessment
	Management of radioactive waste from a major core damage in a BWR power plant
	Some studies related to decommissioning of nuclear reactors
	Kvalitetssäkring av transportbehållare för radioaktivt material
	Geologifrågor i samband med slutförvar av kärnbränsle i det fennoskandiska urberget
	Optimization of radiation protection at nuclear power plants
	Principles for decisions involving environmental and health risks
	Risk analysis and safety rationale
	Prevention of catastrophic failure in pressure vessels and pipings
	Crack arrest - additional safety against catastrophic fracture
	Nordiska transporter
	Intergranular stress corrosion cracking
	Severe accidents analysis - a Nordic study of codes
	Corrosion in seawater systems
	Training diagnostic skills for nuclear power plants
	Nordic Reactor Safety Research 1981 - 85. Summary report of the NKA/SÄK Programme
	Sjøvannsbestandige materialer - kunnskapsoverføring fra kjernekraft til offshoreindustrien
	Risk analysis uses and techniques in the non-nuclear field - a Nordic perspective
	Emnesorienterte prosjektkataloger - en katalysator for nordisk samarbeid?
	The human component in the safety of complex systems
	Management of radioactive waste resulting from nuclear fuel damage
	Organization for safety
	Long-term properties of bituminized waste products
	Computer codes for small-break loss-of-coolant accidents - a Nordic assessment
	Corrosion in the nuclear industry - a Nordic survey
	The design process and the use of computerized tools in control room design
	Computer aided operation of complex systems - experimental testing and evaluation
	Human errors in test and maintenance in nuclear power plants - Nordic project work
	Towards more realistic assessment of reactor accident consequences - A Nordic project
	PRA uses and techniques - a Nordic perspective
	Heat transfer correlations in nuclear reactor safety calculations - Vol. II
	Heat transfer correlations in nuclear reactor safety calculations - Vol. I
	Enlarged Nordic Cooperative Program On Nuclear Safety. Nuclide content in reactor waste
	Nuclide content in reactor waste
	Enlarged Nordic Cooperative Program On Nuclear Safety. Nordic study on reactor waste. Technical Part III Product Characteristics
	Enlarged Nordic Cooperative Program On Nuclear Safety. Nordic study on reactor waste. Technical Part I Reference System & Technical Part II Safety Analysis
	Enlarged Nordic Cooperative Program On Nuclear Safety. Nordic study on reactor waste. Main report
	Nordic study on reactor waste - Tecnical part I, Reference system - Technical part II, Safety analysis
	Nordic study on reactor waste - Technical part III, Product Characteristics
	Nordic study on reactor waste - Main report