LWR Systems Research Articles

Scientific Motivations for a Reassessment of the Neutron Capture Cross Sections of Erbium Isotopes in the High-Sensitivity Thermal Energy Range for Lwr Systems

Scientific Motivations for a Reassessment of the Neutron Capture Cross Sections of Erbium Isotopes in the High-Sensitivity Thermal Energy Range for Lwr Systems

Silver as a capturing material for iodine released from lead\u2013bismuth eutectic in various conditions

The usage of silver as a filtering material for removal of iodine from the gas phase of a lead–bismuth eutectic based nuclear reactor was investigated in various atmospheres representing normal operation as well as accident conditions. Thermochromatography experiments were performed to quantify the retention experienced on a silver surface by iodine species evaporated from a lead–bismuth eutectic sample. Measured adsorption enthalpies ranged from −171 to − 208 kJ mol−1 with observed differences attributed to various surface effects rather than a change in iodine speciation. The postulated adsorption mechanism is chemisorption of iodine atoms on the silver surface. Metallic silver fulfills the desired criteria for a capturing material in water-free filtering systems to be used as an alternative to traditional alkaline scrubbers commonly used in LWR systems.

Methodology for neutronic uncertainty propagation and application to a UAM-LWR benchmark

This work covers an important point of the benchmark released by the expert group on Uncertainty Analysis in Modeling of Light Water Reactors (UAM-LWR). This ambitious benchmark aims to determine the uncertainty in LWR systems and processes in all stages of calculation, with emphasis on multi-physics (coupled) and multi-scale simulations. Specifically, in this work, a simplified BWR core is used to propagate the uncertainty of nuclear data. Due to a high computational cost in the analysis all fuel assemblies are modeled as fresh. The propagation is subdivided into two levels i) assembly level –with SCALE6.2.1 and SAMPLER module–, and ii) core level, –with PARCSv3.2 and DAKOTA 6.3–. The first level takes into account the uncertainties contained in the master library ENDF/B-VII.1 and as a result a problem-dependent neutronic library in NEMTAB format is obtained using TXT2NTAB code. This is a friendly Matlab code developed within this work. Finally, the uncertainty contained in the neutronic library is further propagated through PARCS. A different approach is presented in this work to propagate the uncertainty between codes. Following this approach only two neutronic libraries are generated, one with the average responses and the other with their standard deviations. Then, the standard deviation and a matrix of perturbation factors are used to perturb the main neutronic parameters. A parallel work is done to propagate the thermal-hydraulic parameters in a PWR core (Mesado et al., ).

Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2 /PuO2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs)-clad systems, particularly for current and near-term build LWRs. R&D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN) and uranium silicide (U3 Si2 ). Candidate cladding materials include advanced stainless steel (FeCrAl) and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R&D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I2 S-LWR), a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I2 S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I2 S-LWR design are U3 Si2 and advanced stainless steel respectively. In addition, the I2 S-LWR design adopts an integral configuration and a fully passive decay heat removal system to provide indefinite cooling capability for a class of accidents. This paper presents the equilibrium cycle core design and reactor physics behaviour of the I2 S-LWR with U3 Si2 and the advanced steel cladding. The results were obtained using the traditional two-stage approach, in which homogenized macroscopic cross-section sets were generated by WIMS and applied in a full 3D core solution with PANTHER. The results obtained with WIMS/PANTHER were compared against the Monte Carlo Serpent code developed by VTT and previously reported results for the I2 S-LWR. The results were found to be in a good agreement (e.g. <200 pcm in reactivity) among the compared codes, giving confidence that the WIMS/PANTHER reactor physics package can be reliably used in modelling advanced LWR systems.

Influence of void fraction on BWR spent fuel direct recycling scenario

Preliminary study on influence of changing void fraction (VF) on SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario for boiling water reactor (BWR) spent fuel direct recycling scheme has been carried out. Several VF values of BWR have been investigated to determine the criticality of reactor. The VF values range from 20% to 60%. The fraction of spent fuel to the total loaded fuel was changed from 5% to 20%. The required uranium enrichment for criticality becomes higher with the increasing of VF as well as the enlarging of the fraction of spent fuel in loaded fuel. The neutron spectra become harder with the augmenting of VF. The plutonium and minor actinides isotopes are produced more in the reactor.

Thorium breeder and burner fuel cycles in reduced-moderation LWRs compared to fast reactors

In this paper, a reduced-moderation LWR (RMLWR) system is compared to a sodium-cooled fast reactor (SFR) as a reference fast system for sustained transuranic burning and self-sufficient break-even cycles using thorium (Th) as the fertile material. SFRs can achieve a higher discharge burn-up (and incineration rate – in the case of burners) than RMLWRs, which results in a highly significant reduction of ∼40% and ∼60% in fuel reprocessing and fabrication requirements for breeder and burner reactors respectively. With sufficiently reduced moderation, the repository radiotoxicity and decay heat of actinide process waste from RMLWRs is very similar to that from SFRs, assuming same fraction of process losses per reprocessed fuel inventory, for the breeder and burner cases. In all cases, fuel fabrication is challenging due to high energy gamma and spontaneous neutron sources. The gamma source is comparable in the burner cores but the spontaneous neutron source in the burner options is approximately an order of magnitude lower for the SFR after a large number of recycles. Th fuel is highly preferable to uranium fuel in RMLWRs due to a large improvement in the void coefficient, which must be kept negative. This constraint makes it difficult, if not impossible, to fuel an RMLWR burner or breeder with uranium. Similarly, the void coefficient in SFRs is significantly better with Th fuel, although it is probably not essential for this to be negative. Although the neutron spectrum and core geometry are substantially different, RMLWRs are a variant of conventional LWR technology. RMLWRs have never been built, whereas some SFRs have been built and operated. Despite this, SFRs are anticipated to have a higher capital cost, which appears to outweigh the lower reprocessing and fuel fabrication requirements, although this conclusion is sensitive to the reactor size, cost and availability factor assumptions.

SUPEL Scenario for PWR Spent Fuel Direct Recycling Scheme

Study on SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario for PWR spent fuel direct recycling scheme has been performed. Several spent PWR fuel compositions in loaded fuel has been investigated to achive the criticality of reactor. The reactor can obtain it criticality for 4.5 a% of UO2 enrichment with at maximum 8.0 a% of spent fuel fraction in loaded fuel. The neutron spectra become harder with the raising of UO2 enrichment in the loaded fresh fuel as well as the increasing of the fraction of spent fuel in the core.

Subgroup decomposition method

This paper presents a new method which is an improvement over the consistent generalized energy condensation theory in which the energy dependence of the angular flux and cross sections is expanded in continuous or discrete orthogonal basis functions. The method extends the cross section condensation process, preserving spectral accuracy in condensed-group transport calculations in a simpler and more direct manner, without the need for this expansion in energy. The new “group decomposition” method directly couples a consistent coarse-group criticality calculation with a set of fixed-source “decomposition sweeps” to obtain the fine-group spectrum without the need to solve for higher energy moments of the flux. The method is derived in general geometry and verified with a 1D discrete ordinates lattice cell problem. In addition, one significant application of the method in whole-core reactor analysis, cross section recondensation, is explored in detail and shown to be highly effective at correcting the cross sections on-the-fly for spectral core environment effects. This is demonstrated with two 1D benchmark problems characteristic of LWR and VHTR systems.

ICONE19-43187 Development of real time sensor for under sodium viewer

For the commercialization of the sodium cooled FBR system, it is desirable to develop a plant concept that has economic competitiveness compatible with the LWR system and excels in safety. Since it is difficult to perform inspection and maintenance of the core internals at FBR, such as the reactor vessel inner wall and the core support structure, measures have been taken to ensure safety and reliability by allowing a sufficiently large margin in the design reducing the burden of inspection and maintenance. To date due to problems such as the effect of radiation penetrating from the reactor core, the opacity and high chemical activity of sodium, and the fact that the inspection of reactor vessel of the FBR is performed at a high temperature, approximately 200 degree C, to avoid freezing of sodium, no method of inspecting structures inside the reactor vessel has been developed. In this study, aims to experimentally produce a real-time sensor which is used for the underwater and under-sodium test to demonstrate their applicability to the FBR system. The real-time sensor would be a piezoelectric element type sensor with a resolution of approximately 2.0mm, a resolution of sufficient quality to allow for the identification of deformation, the failure or dropping of components, and an image processing time per one image of approximately 0.5 second. Concerning the backing material, aluminum titanate (alumatite), wollastonite and hexagonal boron nitride were selected for the first screening. Based on the test result, alumatite that has an excellent damping property was finally selected for the trial production of the sensor element. We have designed and made 256 channel matrix-arrayed sensor and high speed processing system. It was shown that the real-time sensor was achieved about 2.0mm resolution and about 0.5 seconds/frame signal processing time, and succeeded taking the first under-sodium movie of the world.

軽水炉増出力システム・機器の要素技術開発(3)

軽水炉増出力システム・機器の要素技術開発(3)

軽水炉増出力システム・機器の要素技術開発(1)

軽水炉増出力システム・機器の要素技術開発(1)

軽水炉増出力システム・機器の要素技術開発（２）

軽水炉増出力システム・機器の要素技術開発（２）

Radical innovation for a second nuclear era : Conditions for success

Is it possible to create the conditions for a nuclear revival through suitable technological response which would win over public opinions, industry and governments in the future? This communication analyses the specific characteristics of the nuclear technology which create and maintain a de facto lock-in around the LWR, the conditions for success of the initial technology push, and the factors which made nuclear technology unsuited to its changing environment (problems of social acceptance, competition of new power generation techniques, liberalisation of power industries). The aim is to identify the respective possibility either of pursuing the trajectory of incremental innovations, or of breaking loose from the technological path dependency of LWR systems with a radical innovation such as the Accelerator-Driven System (ADS), in order to overcome the problem of unsuitability of nuclear technology and to respond to the energy and environmental challenges of the XXI e century. Such a radical innovation has to meet numerous safety, non-proliferation and economic criteria, but it could only be part of the response to the social and economic existing constraints on the diffusion of the nuclear technology. How to organise the support to its development without the certainty of its success.

PWR life evaluation project of edf

The massive contribution to power generation of its recent PWR has provided the Electricite de France with a strong incentive to carefully examine the potential for the safe and economical service life of these units. Due to their low fuel cost, and to the lack of likely competitors, LWR systems should be ensured of a long future, by comparison with previous thermal facilities. Moreover, the moderate level of thermal performances, as well as the sophisticated design methods applied to the main components give them the best inherent chances of endurance. In 1986 EDF undertook an important programme of work on this subject. In the first phase, all available elements are gathered and reviewed, from design, fabrication, in-service experience and surveillance. Thereafter, complementary studies will be made as far as necessary to improve the knowledge and the control of aging processes, and optimize the maintenance and operation practices. Preliminary conclusions can be drawn from the ongoing studies. Favourable statements can be made regarding the fatigue resistance of the main primary circuits, the irradiation aging of the RPV and the basic behaviours of the main containment structures. In spite of some concern with respect to SG tubing corrosion, which could lead to higher maintenance and repair costs higher than wished for the oldest units, there is no doubt that most of the units will be able to supply a service life of at least 40 years and probably longer.

Geyser, a simple, new heating reactor of high inherent safety

Reactors for district heating are characterized by low power density and extremely slow load variation. Based on these requirements a completely new system GEYSER is under development at SIN. The GEYSER-concept represents a breakthrough in this area of nuclear technology. For the very first time an entirely passive LWR system with inherently safe operation is proposed. Normal unattended operation can be guaranteed without any of the usual active components. High reliability is achieved without cost intensive redundancies and corresponding active control systems. Sensitivity to external an internal events is drastically reduced at a very low cost, at the same time human influences are completely eliminated.

Strain-induced corrosion cracking of low-alloy steels in LWR systems — case histories and identification of conditions leading to susceptibility

The term “strain-induced corrosion cracking” (SICC) is introduced to describe crack formation involving dynamic straining, but in the absence of obvious, cyclic loading. Its origins in slow-strain-rate testing and in corrosion failures in boiler systems are described and the links with “classical” stress corrosion cracking and low-cycle corrosion fatigue are identified. Four areas, in which SICC of low-alloy steels in LWR systems has occurred, are described in detail and the typical features are used, together with literature data from laboratory testing, to identify conditions leading to susceptibility. Indications are given of remedial measures and of areas in which further work is necessary.

The Industry/EPRI Advanced Light Water Reactor Program

For the United States nuclear power industry to remain viable, it must be prepared to meet the expected need for new generating capacity in the late 1990s with an improved reactor system. The best hope of meeting this requirement is with evolutionary changes in current LWR systems through system simplification and reevaluation of safety and operational design margins. The grid characteristics and the difficulty in raising capital for large projects indicate that smaller light water reactors (400 to 600 MWe) may play an important role the next generation.

Aging of cast duplex stainless steels in LWR systems

A program is being conducted to investigate the significance of in-service embrittlement of cast duplex stainless steels under light-water reactor operating conditions. The existing data are evaluated to determine the expected embrittlement of cast components during the operating lifetime of reactors and to define the objectives and scope of the investigation. This presentation describes the status of the program. Data for the metallurgical characterization of the various cast stainless steels used in the investigation are presented. Charpy impact tests on short-term aged material indicate that CF-3 stainless steels are less susceptible to embrittlement than CF-8 or CF-8M stainless steels. Microstructural characterization of cast stainless steels that were obtained from Georg Fischer Co. and aged for up to 70000 h at 300, 350, and 400°C reveals the formation of four different types of precipitates that are not α'. Embrittlement of the ferrite phase is primarily due to pinning of the dislocations by two of these precipitates, designated as Type M and Type X. The ferrite phase is embrittled after ≈ 8 y at 300°C and shows cleavage fracture. Examination of the fracture surfaces of the impact-test specimens indicates that the toughness of the long-term aged material is determined by the austenite phase.

Preliminary design of a containment to withstand core melt for a 1300 MWe LWR system

This paper identifies several of the possible accident sequences including postulated reactor core melt of such low probability that typical light water reactor nuclear power stations are not specifically designed to mitigate their affects. It then identifies a conceptual design and presents a preliminary cost study for a passive containment system designed to mitigate the consequences of such low probability accident sequences. The conceptual containment design thus developed appears to offer an economic alternative to underground siting or more elaborate and redundant emergency core cooling systems.

Underground siting of nuclear power plants with emphasis on the ‘cut-and-cover’ technique

This paper provides a survey of activities in the field of underground siting in Europe. Emphasis is placed on the ‘cut-and-cover’ technique because of the geological and topographical conditions prevailing in Germany. Basic considerations show the potential of an additional level of containment, of better protection against external effects, and of better conditions for decommissioning. Design concepts for HTR systems and especially for LWR systems which do not require extensive changes in the above-ground design are defined, and a preliminary assessment is given.