Simplified criteria for a comparison of the accidental behaviour of Gen IV nuclear reactors and of PWRS (original) (raw)
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Fusion Technology, 1992
Radioactivity induced in a typical fusion power reactor was calculated for all elements with atomic number Z < 84 and for different irradiation times. It was shown that the shutdown activity varies widely with the duration of the irradiation time. In general, the activity induced by radionuclides with halflives that are significantly longer than the period of irradiation increases with increasing the irradiation time. On the other hand, the level of activity generated by any radionuclide with a half-life which is significantly shorter than the reactor lifetime reaches a peak early during irradiation and then may starts to drop to lower value before the end of irradiation. The severity of this peaking is determined by the destruction rate of the parent element The activities generated by long-lived nuclides (important for waste management) in any fusion reactor with life time in the order of 30 years reach their peak values at end-of-life. In the mean time, using the activity and decay heat values generated by short and intermediate-lived radionuclides at the end of reactor life to represent the worst case values used in safety analyses related to a loss of coolant accident (LOCA) and accidental release of radioactive inventory might lead to a substantial underestimation of the results.
Inherent safety concepts in nuclear power reactors
Sadhana, 1989
Different inherent safety concepts being considered in fast and thermal reactors are presented after outlining the basic goals of nuclear reactor safety, the 'defence in depth' philosophy to achieve these goals and the characteristics affecting the safety of liquid metal fast breeder and light water reactors. The inherent safety potential of fast reactors with respect to different sizes and types of fuel is also discussed. Finally, the approach proposed for the Indian Prototype Fast Breeder Reactor (PFBR), which is in the detailed design stage, is also presented.
This section contains the description of related engineering and analytical processes that are used generally in nuclear engineering related to the design and operation of nuclear processes. Chapters 19 and 20 describe the safety evaluations that are used for nuclear facilities. Chapter 19 introduces the risk assessment and safety analysis process that is used for nuclear reactors that are licensed in the United States by the Nuclear Regulatory Commission (NRC). This process has evolved from a relatively simple safety analysis used in the 1950s to a detailed risk assessment process that is used today. Chapter 20 describes the process used in the United States by the Department of Energy for safety analysis of its facilities. It is more prescriptive and less probability and risk based than the process used by the NRC.
The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants
Science and Technology of Nuclear Installations, 2012
Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence) to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP). After a first project in the 6th Framework Programme (FP6) of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments) by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…), source term issues (mainly iodine behaviour). The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation res...
2008
An accident, thermal fluids, and reactor physics phenomena identification and ranking process was conducted by a panel of experts on the next generation nuclear plant (NGNP) design (consideration given to both pebble-bed and prismatic gas-cooled reactor configurations). Safety-relevant phenomena, importance, and knowledge base were assessed for the following event classes: 1. normal operation (including some reactor physics aspects), 2. general loss of forced circulation (G-LOFC), 3. pressurized loss-of-forced circulation (P-LOFC), 4. depressurized loss-of-forced circulation (D-LOFC), 5. air ingress (following D-LOFC), 6. reactivity transients-including anticipated transients without scram (ATWS), 7. processes coupled via intermediate heat exchanger (IHX) (IHX failure with molten salt), and 8. steam/water ingress. The panel's judgment of the importance ranking of a given phenomenon (or process) was based on the effect it had on one or more figures of merit or evaluation criteria. These included public and worker dose, fuel failure, and primary (and other safety) system integrity. The major phenomena of concern that were identified and categorized as high importance combined with medium to low knowledge follow: • core coolant bypass flows (normal operation), • power/flux profiles (normal operation), • outlet plenum flows (normal operation), • reactivity-temperature feedback coefficients for high-plutonium-content cores (normal operation and accidents), vi The PIRT process for the NGNP completes a major step towards assessing NRC's research and development needs necessary to support its licensing activities, and the reports satisfy a major EPAct milestone. The results will be used by the agency to: (1) prioritize NRC's confirmatory research activities to address the safety-significant NGNP issues, (2) inform decisions regarding the development of independent and confirmatory analytical tools for safety analysis, (3) assist in defining test data needs for the validation and verification of analytical tools and codes, and (4) provide insights for the review of vendors' safety analysis and supporting data bases.
Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept
Sustainability, 2012
Since a fast reactor core with uranium-plutonium fuel is not in its most reactive configuration under operating conditions, redistribution of the core materials (fuel, steel, sodium) during a core disruptive accident (CDA) may lead to recriticalities and as a consequence to severe nuclear power excursions. The prevention, or at least the mitigation, of core disruption is therefore of the utmost importance. In the current paper, we analyze an innovative fast reactor concept developed within the CP-ESFR European project, focusing on the phenomena affecting the initiation and the transition phases of an unprotected loss of flow (ULOF) accident. Key phenomena for the initiation phase are coolant boiling onset and further voiding of the core that lead to a reactivity increase in the case of a positive void reactivity effect. Therefore, the first level of optimization involves the reduction, by design, of the positive void effect in order to avoid entering a severe accident. If the core disruption cannot be avoided, the accident enters into the transition phase, characterized by the progression of core melting and recriticalities due to fuel compaction. Dedicated features that enhance and guarantee a sufficient and timely fuel discharge are considered for the optimization of this phase.
Nuclear Engineering and Design, 2000
As part of the Combustion Améliorée du Plutonium dans les Réacteurs Avancés/Consommation D'Actinides et de Déchets dans les Réacteurs Avancés (CAPRA/CADRA) program the feasibility of reactor systems with different neutron spectra and coolants is investigated to burn plutonium and also to destruct minor actinides and long lived fission products. In this paper, we deal with reactor cores with fast spectrum and metal cooling. The design of this type of CAPRA/CADRA cores shows significant differences compared e.g. to conventional fast reactor cores. The high Pu-enrichment and the high minor actinide load have an important influence on the core meltdown behavior and the associated recriticality risk. To cope with this risk, inherent design features and special measures/devices are investigated for their potential of early fuel discharge to reduce the criticality of the reactor core. An assessment of such measures/devices, which could provide an additional line of defense against severe accident development, is given. Within the CAPRA/CADRA program, also accelerator driven subcritical systems are investigated for performing the task of transmutation and incineration. In these fast neutron systems with a strong external neutron source, the kinetic behavior is different to a critical core and new strategies and measures for accident prevention have to be investigated.
Nuclear Engineering and Design, 2019
One of the main research goals of the GEN-IV systems is enhancing their safety compared with the former Sodium-Cooled Fast Reactor (SFR) designs. A key issue is the capability of accidents prevention as well as of demonstrating that their consequences do not violate the safety criteria. In order to fulfill such requirements, risk analyses of severe core disruptive accidents are performed. Since the beginning of the SFR development, Hypothetical Core Disruptive Accidents (HCDAs) have played an outstanding role. Numerous safety analyses have been performed for developing and licensing past SFR designs and nowadays a large database of results is available. In particular, a large amount of results of the mechanistic SIMMER-II and SIMMER-III/IV analyses for various core designs and different power classes is available at the Karlsruhe Institute of Technology (KIT). The current paper describes the probabilistic approach based on the Phenomenological Relationship Diagram (PRD), which is used to evaluate the Probability Distribution Function (PDF) of the thermal energy release during the transition phase of an unprotected loss of flow accident scenario for a SFR. The technique allows taking into account the mechanistic nature of the accident scenario. In fact, the available results of the mechanistic analyses of HCDAs in SFRs are used to assess the PDFs of the dominant phenomena affecting the thermal energy release, which are propagated in the PRD by employing a Monte Carlo method.
The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radio...