Um dos principais mecanismos de falha que causam riscos de fratura a reatores de água pressurizada é a corrosão sob tensão de ligas metálicas em água do circuito primário (CSTAP). É causada por uma combinação das tensões de tração, meio ambiente em temperatura e microestruturas metalúrgicas susceptíveis. Ela pode ocorrer, dentre outros locais, nos bocais do mecanismo de acionamento das barras de controle. Essa fratura pode causar acidentes que comprometem a segurança nuclear através do bloqueio das barras de controle e vazamentos de água do circuito primário reduzindo a confiabilidade e a vida útil do reator. O objetivo desta Tese de Doutorado é o estudo de modelos e uma proposta de modelagem para fraturas por corrosão sob tensão em liga 75Ni15Cr9Fe (liga 600), em água de circuito primário de reator de água pressurizada nesses bocais. São superpostos modelos eletroquímicos e de mecânica da fratura e validados com dados obtidos em experimentos e na literatura. Na parte experimental foram utilizados resultados obtidos pelo CDTN no equipamento recém-instalado de ensaio por taxa de deformação lenta. Na literatura está proposto um diagrama que exprime a condição termodinâmica de ocorrerem diversos modos de CSTAP na liga 600: partiu-se de diagramas de potencial x pH (diagramas de Pourbaix)...
Analysts and decision makers frequently want estimates of the cost of technologies that have yet to be developed or deployed. Small modular reactors (SMRs), which could become part of a portfolio of carbon-free energy sources, are one such technology. Existing estimates of likely SMR costs rely on problematic top-down approaches or bottom-up assessments that are proprietary. When done properly, expert elicitations can complement these approaches. We developed detailed technical descriptions of two SMR designs and then conduced elicitation interviews in which we obtained probabilistic judgments from 16 experts who are involved in, or have access to, engineering-economic assessments of SMR projects. Here, we report estimates of the overnight cost and construction duration for five reactor-deployment scenarios that involve a large reactor and two light water SMRs. Consistent with the uncertainty introduced by past cost overruns and construction delays, median estimates of the cost of new large plants vary by more than a factor of 2.5. Expert judgments about likely SMR costs display an even wider range. Median estimates for a 45 megawatts-electric (MWe) SMR range from $4,000 to $16,300/kWe and from $3,200 to $7,100/kWe for a 225-MWe SMR. Sources of disagreement are highlighted...
The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to be isolated from the environment for thousands of years. In addition, plutonium and other actinides, after the decay of fission products, could become targets for weapon proliferators. Furthermore, only a small fraction of the energy potential in the fuel is being used. All these concerns can be addressed if a closed fuel cycle strategy is considered offering the possibility for partitioning and transmutation of long lived radioactive waste, enhanced proliferation resistance, and improved utilization of natural resources. It is generally believed that dedicated advanced reactor systems have to be designed in order to perform the task of nuclear waste transmutation effectively. The development and deployment of such innovative systems is technically and economically challenging. In this thesis, a possibility of constraining the generation of long lived radioactive waste through multi-recycling of Trans-uranic actinides (TRU) in existing Light Water Reactors (LWR has been studied. Thorium based and fertile free fuels (FFF) were analyzed as the most attractive candidates for TRU burning in LWRs. Although both fuel types can destroy TRU at comparable rates (about 1150 kg/GWe-Year in FFF and up to 900 kg/GWe-Year in Th) and achieve comparable fractional TRU burnup (close to 50a/o)...
by John Edward Kelly.; Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1980.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE; Includes bibliographical references.
by Kord Sterling Smith.; Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1980.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE.; Vita.; Includes bibliographical references.
by John Edward Rivera.; Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981.; MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE.; Includes bibliographical references.
The nature and characteristics of nuclear reactor transients
induced by control rod motions are important to light-water
reactor safety analyses. Rod motion influences both local
neutron absorption rates and the local neutron spectra.
Studies on specific systems suggest that accurate prediction
of transients requires that both the absorption rate change
and the spectral change are necessary to represent control rod
motion.; New England Electric System, Northeast Utilities Service Co. under the M.I.T. Energy Electric Power Program
This report summarizes a two-year effort by the M.I.T. Light Water
Reactor Study Group to assess the institutional, regulatory, technical, and
economic factors influencing the development and deployment of LWR technology.
The nuclear industry is confronted by a mix of problems which, if not
addressed, may soon eliminate LWRs as a practical source of electric energy.
The Study Group found that technical developments could improve nuclear plant
capacity factors by 10 percent; furthermore, substantial economic benefits are
possible through better use of existing technology, further technological
improvements, and various financing schemes. However, the most pronounced
problems are institutional and social, not technical and economic. Regulatory
and institutional problems in licensing, constructing, and operating nuclear
plants have created such uncertainty in the electric utility sector that the
economic and environmental advantages of LWRs are seriously jeopardized.
Regulatory constraints, unpredictability of government policy, unnecessary
construction delays, and the resultant difficulty in obtaining the large-scale
financing needed for new plant construction all discourage the electric
utility sector from making long-term commitments to nuclear power. In the
absence of a concerted government attempt to resolve these and other problems...
To increase the power density and maximum allowable fuel burnup in light water reactors, new fuel rod designs are investigated. Such fuel is desirable for improving the economic performance light water reactors loaded with transuranic-bearing fuel for transmutation, as well as those using UO2 fuel. A proposal for using silicon carbide duplex as fuel cladding is investigated. The cladding consists of a monolithic inner layer surrounded by a tightly wound fiber-matrix composite. The monolith layer retains the volatile fission products while the composite adds strength. The FRAPCON steady-state thermo-mechanical fuel rod modeling code is used to examine the performance of SiC cladding at high fuel burnup and high power density. Empirical models are developed to describe the physical properties of the composite as a function of operating temperature and neutron fluence. A comparison of the behavior of the SiC cladding to the conventional Zircaloy cladding demonstrates that the SiC has superior resistance to creep and mechanical degradation due to radiation or oxidation. However, the lower thermal conductivity of the SiC is a major issue, which results in significantly increased peak fuel temperatures. Mixed U02-PuO2 fuel is also examined in place of traditional UO2 pellets...
This work investigates whether the advanced light water reactor designs with passive safety systems are more desirable than advanced reactor designs with active safety systems from the point of view of uncertainty in the performance of safety systems as well as the economic implications of the passive safety systems. Two advanced pressurized water reactors and two advanced boiling water reactors, one representing passive reactors and the other active reactors for each type of coolant, are compared in terms of operation and responses to accidents as reported by the vendors. Considering a simplified decay heat removal system that utilizes an isolation condenser for decay heat removal, the uncertainty in the main parameters affecting the system performance upon a reactor isolation accident is characterized when the system is to rely on natural convection and when it is to rely on a pump to remove the core heat. It is found that the passive system is less certain in its performance if the pump of the active system is tested at least once every five months. In addition, a cost model is used to evaluate the economic differences and benefits between the active and passive reactors. It is found that while the passive systems could have the benefit of fewer components to inspect and maintain during operation...
Nanofluids at very low concentrations experimentally exhibit a substantial increase in Critical Heat Flux (CHF) compared to water. The use of a nanofluid in the In-Vessel Retention (IVR) severe accident management strategy, employed in Advanced Light Water Reactors, was investigated. A model simulating the two-phase flow and heat transfer on the reactor vessel outer surface quantified the increase in decay power that can be removed using a nanofluid, predicting that the use of a nanofluid will allow a stable operating power ~40% greater than the power allowable using water to be achieved, while holding the Minimum Departure from Nucleate Boiling Ratio (MDNBR) constant. A nanofluid injection system that would take advantage of the enhanced CHF properties of the nanofluid in order to provide a higher safety margin than the current IVR strategy or, for given margin, enable IVR at higher core power, is proposed. A risk-informed analysis has revealed that this injection system has a reasonably high success probability of 0.99, comparable to the success probability without the injection system. Potential regulatory, environmental, and health risk issues were analyzed, and it was concluded that the current regulatory regimes are adequate for ensuring that the implementation of nanofluids in IVR will not endanger public health and safety. However...
The economic advantages of longer fuel cycle, improved fuel utilization and reduced spent fuel storage have been driving the nuclear industry to pursue higher discharge burnup of Light Water Reactor (LWR) fuel. A design basis accident, the Reactivity Initiated Accident (RIA), became a concern for further increase of burnup as simulated RIA tests revealed lower enthalpy threshold for fuel failure associated with fuel dispersal, which may compromise the core coolability and/or cause radiological release should this happened in LWRs. Valuable information on the behavior of high burnup fuel during RIA are provided by the simulation tests. However atypical design and operating conditions in simulated tests limited the application of experimental data directly to evaluate the failure potential of LWR fuels. To better interpret the experimental results and improve the capability of the fuel performance codes to predict high burnup fuel behavior, this thesis developed mechanistic models of high burnup fuel during an RIA and implemented models in a transient fuel performance code FRAPTRAN 1.3. Fission gas release (FGR) and swelling were systematically modeled to quantify gaseous loading effects. The grain boundary fission gas inventory is simulated prior to the transient using a diffusion model in FRAPCON 3.3 code. The restructuring of high burnup fuel in rim region is described in terms of porosity...
Fonte: KOREAN NUCLEAR SOCPublicador: KOREAN NUCLEAR SOC
Tipo: Articles in JournalsFormato: Printed
Relevância na Pesquisa
The past two decades were mainly devoted to model validation and computer code verification against global corium experiments, code application to reactor situations, and investigation of the role of melt properties in steam explosion energetics. Corium data were essentially provided by JRC-Ispra in the FARO and KROTOS facilities and by KAERI in the TROI facility. Verification of code applicability to reactor situations was performed essentially in the frame of the international OECD/SERENA programme. The paper makes a synthesis of the findings made during the above-mentioned period and expresses a personal view of the author with respect to the progress made and expected for the resolution of the steam explosion issue for light water reactors.; JRC.F.4-Safety of future nuclear reactors
NILSSON KARL-FREDRIK; TAYLOR NIGEL; DAHLBERG M.; FAIDY C.; WILKE U; CHAPULIOT S; KALKHOF D; BRETHERTON I; CHURCH M; SOLIN J; CATALANO J
Fonte: European CommissionPublicador: European Commission
Tipo: BooksFormato: Printed
Relevância na Pesquisa
In nuclear plant piping systems thermal fatigue damage can arise at locations where there is turbulent mixing of different temperature flows. The severity of this phenomenon can be difficult to assess via plant instrumentation due to the high frequencies involved. In Europe the existing approaches to high cycle thermal fatigue have been successful in providing margins against pipe ruptures. Nonetheless there have been instances of thermal fatigue damage and over the last 10 years several recent R&D programmes have been devoted to developing better understanding of the induced thermal loads and associated damage mechanisms. To exploit this work, in 2003 the Network for Evaluation of Structural Components (NESC) set up a project involving both utilities and R&D organizations. Its aim was to produce a consensus methodology for assessing high cycle thermal fatigue in piping components, with special attention to turbulent mixing phenomena at mixing tees in light water reactor systems. It has involved the collaboration of over 10 organisations from 5 European countries. All have participated on an 100% in-kind basis. Wherever possible, advantage has been taken of recent R&D work at national, European and international levels. The work programme focused on two main aspects:
a) creating a database of service and mock-up data for better understanding thermal fatigue damage mechanisms and for validating the procedure.
b) developing a European multi-level thermal fatigue damage procedure...
The Reactor Safety Study (RSS) or Wash-1400 developed a
methodology estimating the public risk from light water nuclear
reactors. In order to give further insights into this study,
a sensitivity analysis has been performed to determine the
significant contributors to risk for both the PWR and BWR.
The sensitivity to variation of the point values of the failure
probabilities reported in the RSS was determined for the
safety systems identified therein, as well as for many of the
generic classes from which individual failures contributed to
system failures. Increasing as well as decreasing point values
were considered. An analysis of the sensitivity to increasing
uncertainty in system failure probabilities was also performed.
The sensitivity parameters chosen were release category prob-
abilities, core melt probability, and the risk parameters of
early fatalities, latent cancers and total property damage.
The latter three are adequate for describing all public risks
identified in the RSS. The results indicate reductions of
public risk by less than a factor of two for factor reductions
in system or generic failure probabilities as hignh as one hundred.
There also appears to be more benefit in monitoring the most
sensitive systems to verify adherence to RSS failure rates
than to backfitting present reactors. The sensitivity analysis
results do indicate...
The feasibility of using various uranium-free fuels for plutonium incineration in present light water reactors is investigated. Two major categories of inert matrix fuels are studied: composite ceramic fuel particles dispersed in another ceramic matrix (CERCER) and ceramic fuel particles dispersed into a metallic matrix (CERMET). In the category of CERCER, the current world wide research effort has been focused on three matrix candidates: (1) Spinel (MgAl2O4); (2) CeO2, and (3) MgO. In contrast, there are still no emerging commonly accepted matrix candidates for a CERMET. The fuel may consist of plutonium, minor actinides (MA), or both which are termed trans-uranium (TRU) fuel. The transmutation rate and the transmuted fraction of initial loadings are calculated using CASMO-4. Different inert matrix fuels have similar burning abilities in terms of how much and how fast the Pu, MA or TRU can be burned, and they are all superior to the mixed UO2-PuO2 (MOX) fuel. From this point of view, there is no good reason to favor one inert matrix over another. The burning rates in terms of kg/(GWe-Year) of different inert matrix fuels are quite stable with regard to changing the moderation level (or H/HM ratio) in the core. Changing initial loadings and changing power densities can not result in large change in the burned percentage of initial loadings and burning rate. Lack of U-238 and the neutronic characteristics of plutonium lead to degradation of safety related kinetic parameters. It is found that various inert matrix fuels have similar values for the Doppler coefficient...
The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current
light water reactors, with a specific focus on pressurized water reactors.
The types of changes which have been examined are: (1) re-optimization of fuel pin diameter and lattice pitch, (2) Axial power shaping by enrichment gradation in fresh fuel, (3) Use of 6-batch cores with semi-annual refueling, (4) Use of 6-batch cores with annual refueling, hence greater extended (.doubled) burnup, (5) Use of radial reflector assemblies, (6) Use of internally heterogeneous cores (simple seed/blanket configurations), (7) Use of power/temperature coastdown at the end of life to extend burnup, (8) Use of metal or diluted oxide fuel, (9) Use of thorium, and (10) Use of isotopically separated low a cladding material.
State-of-the-art LWR computational methods, LEOPARD/PDQ-7/FLARE-G, were used to investigate these modifications. The most effective way found to improve uranium ore utilization is to increase the discharge burnup. Ore savings on the order of 20% can be realized if greatly extended burnup (-
double that of current practice) is combined with an increase in the number of batches in the core from 3 to 6. The major conclusion of this study is that cumulative reductions in ore usage of on the order of 30% are fore-
seeable relative to a current PWR operating on the once-through fuel cycle...
Building a new generation of fission reactors in the United States presents
many technical and regulatory challenges. One important challenge is the need
to share and present results from new high-fidelity, high-performance
simulations in an easily usable way. Since modern multiscale, multi-physics
simulations can generate petabytes of data, they will require the development
of new techniques and methods to reduce the data to familiar quantities of
interest (e.g., pin powers, temperatures) with a more reasonable resolution and
size. Furthermore, some of the results from these simulations may be new
quantities for which visualization and analysis techniques are not immediately
available in the community and need to be developed.
This paper describes a new system for managing high-performance simulation
results in a domain-specific way that naturally exposes quantities of interest
for light water and sodium-cooled fast reactors. It describes requirements to
build such a system and the technical challenges faced in its development at
all levels (simulation, user interface, etc.). An example comparing results
from two different simulation suites for a single assembly in a light-water
reactor is presented, along with a detailed discussion of the system's
requirements and design.; Comment: Article on NiCE's Reactor Analyzer. 23 pages. Keywords: modeling...
Light water reactors (LWRs) are the world?s dominant nuclear reactor system. Uranium (U)-fuelled LWRs produce long-lived transuranic (TRU) isotopes. TRUs can be recycled in LWRs or fast reactors. The thermal neutron spectrum in LWRs is less suitable for burning TRUs as this causes a build-up of TRU isotopes with low fission probability. This increases the fissile feed requirements, which tends to result in a positive void coefficient (VC) and hence the reactor is unsafe to operate. Use of reduced-moderation LWRs can improve TRU transmutation performance, but the VC is still severely limiting for these designs. Reduced-moderation pressurized water reactors (RMPWRs) and boiling water reactors (RBWRs) are considered in this study.
Using thorium (Th) instead of U as the fertile fuel component can greatly improve the VC. However, Th-based transmutation is a much less developed technology than U-based transmutation. In this thesis, the feasibility and fuel cycle performance of full TRU recycle in Th-fuelled RMPWRs and RBWRs are evaluated. Neutronic performance is greatly improved by spatial separation of TRU and 233-6U, primarily implemented here using heterogeneous RMPWR and RBWR assembly designs.
In a RMPWR, the water to fuel ratio must be reduced to around 50% of the normal value to allow full actinide recycle. If implemented by retrofitting an existing PWR...