SEMER: a simple code for the economic evaluation of nuclear and fossil energy-based power production systems

The code system, SEMER, was recently developed to evaluate the economic impact of various nuclear reactors and associated innovations. Models for nearly all fossil energy-based systems were also included to provide a basis for cost comparisons.Essentially, SEMER includes three types of model libraries: the global model, for a rapid estimation of various nuclear and fossil energy-based systems, the detailed models, for the finer cost evaluation of individual components and circuits in a PWR type of reactor and the fuel cycle models, for PWRS, HTRs and FBRs, allowing the cost estimations related to all the steps in the nuclear fuel cycle, including reprocessing and disposal.This paper summarises our on-going investigations on new developments in, and on the validation of, the SEMER system.Details of the modelling principles, and the results of validation carried out in the context of an EDF/CEA Joint Protocol Agreement, are also presented.First results of this validation are highly encouraging:

• Relative errors for the total kWh or overnight and investment costs are less than 5% for large PWR systems operating in France or other countries.

• These errors are less than 3% for small-sized compact PWRs and they are of the order of 4–7% for HTRs (as compared to IAEA estimations).

• For fossil energy-based power plants, the relative error, even with slightly different cost breakdown between SEMER and that of existing installations, is from 5 to 20%.

• Similarly, errors on the nuclear fuel cycle costs are about 1–4%, compared to published reference values.

Qualification of the LF-eddy current technique for the inspection of stainless steel cladding and applications on the reactor pressure vessel

As part of the re-inspection of the reactor pressure vessel of the nuclear power plant, the low-frequency-eddy current technique was implemented during the 1995 outage. Since then, this inspection technique and the testing equipment have seen steady further development. Therefore, optimization of the entire testing system, including qualification based on the 1995 results, was conducted. The eddy current testing system was designed as a ten-channel test system with sensors having separate transmitter and receiver coils. The first qualification of the testing technique and sensors was performed using a single-channel system; a second qualification was then carried out using the new testing electronics. The sensor design allows for a simultaneous detection of surface and subsurface flaws. This assumes that testing is performed simultaneously using four frequencies. Data analysis and evaluation are performed using a digital multi-frequency regression analysis technique The detection limits determined using this technique led to the definition of the following recording limits for testing in which the required signal-to-noise ratio of 6 dB was reliably observed.

• Detection of surface connected longitudinal and transverse flaws:

• notch, 3 mm deep and 10 mm long, for weave bead cladding;

• notch, 2 mm deep and 20 mm long, for strip weld cladding.

• Detection of embedded planar longitudinal and transverse flaws:

• ligament of 7 mm for 8 mm clad thickness and 3 mm;

• ligament for 4 mm clad thickness, notch starting at the carbon steel base material with a length of 20 mm.

• Detection of embedded volumetric longitudinal and transverse flaws:

• 3 mm diameter side-drilled hole (SDH) for 8 mm clad thickness; ligament, 4 mm. For 4 mm clad thickness: diameter, 2 mm SDH; ligament, 2 mm. All SDHs are 55 mm deep.

PWR steam generator R&D in France: thermal-hydraulic and vibratory aspects

Concomitant with the launching of the French pressurized water reactor (PWR) nuclear power program, a large research and development (R&D) effort was initiated, devoted to the steam generators (SGs). This program, managed cooperatively by Framatome, the SG designer and manufacturer; Electricité de France (EDF), the French electrical utility; and the Commissariat à l'Energie Atomique (CEA), the French Atomic Energy Commission, primarily responsible for nuclear research; was focused on four main objectives:

1. (1) To obtain a better understanding of the physical phenomena existing in these steam generators and leading to SG performance alterations or operating life reductions.

2. (2) To test and validate improved design solutions for the model 51 Framatome steam generator, which was the first one designed under Westinghouse license.

3. (3) To test and validate new Framatome SG designs.

4. (4) To test and validate new, high-performance design tools.

This vast R&D program covers the following theses:

• - SG thermal-hydraulics,

• - SG tube vibration and wear,

• - SG materials (production, corrosion, etc.),

• - Primary and secondary fluid chemistry,

• - SG technology (manufacturing processes, NDT, etc.),

• - SG in-service inspection, and

• - SG maintenance.

These themes are too numerous to be dealt with in a single article. Consequently, the present article will focus on only the first two themes.     

CEA and AREVA R&D on HTR fuel fabrication and presentation of the CAPRI experimental manufacturing line

In the framework of the French V/HTR fuel development and qualification program, the Commissariat à l’Energie Atomique (CEA) and AREVA are conducting R&D projects covering the mastering of UO₂ coated particle and fuel compact fabrication technology. To fulfill this task, a review of past knowledge, of existing technologies and a preliminary laboratory-scale work program have been conducted with the aim of retrieving the know-how on HTR coated particle and compact manufacture:

• The different stages of UO₂ kernel fabrication GSP process have been reviewed, reproduced and improved.

• The experimental conditions for the chemical vapor deposition of coatings have been defined on dummy kernels and development of innovative characterization methods has been carried out.

• Former CERCA compacting process has been reviewed and updated.

In parallel, an experimental manufacturing line for coated particles, named GAIA, and a compacting line based on former CERCA compacting experience have been designed, constructed and are in operation since early 2005 at CEA Cadarache and CERCA Romans, respectively. These two facilities constitute the CAPRI line (CEA and AREVA PRoduction Integrated line).The major objectives of the CAPRI line are:

• to recover and validate past knowledge,

• to produce representative HTR TRISO fuel meeting industrial standards,

• to permit the optimization of reference fabrication processes for kernels and coatings defined previously at a laboratory-scale and the investigation of alternative and innovative fuel design (UCO kernel, ZrC coating),

• to test alternative compact process options and

• to fabricate and characterize fuel required for irradiation and qualification purpose.

This paper presents the status of progress of R&D conducted on HTR fuel particles and compact manufacture by early 2005 and the potential of the laboratory-scale HTR fuel CAPRI line.     

High performance light water reactor

The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:

• A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.

• Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a ‘reference design’, developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the ‘reference design’ was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to ‘calibrate’ the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly.

• Preliminary selection was made for the HPLWR scale, boundary conditions, core and fuel assembly design, reactor pressure vessel, containment, turbine and balance of plant.

• A review of potentially applicable materials for the HPLWR was completed and a preliminary selection of potential in-vessel and ex-vessel candidate materials was made.

• A thorough review of heat transfer at supercritical pressures was completed together with a thermal-hydraulics analysis of potential HPLWR sub-channels. This analytical tool supports the core and fuel assembly design.

• The RELAP5 and the CATHARE 2 codes are being upgraded to supercritical pressures. Thus they can be used to support the HPLWR core design and to perform plant safety analyses.

• Assessment of the HPLWR design constraints, based on current LWR technology was documented. This document stresses the various criteria that must be satisfied in the design (e.g. material, temperature, power, safety criteria, etc.) based on experience gained in the design of PWR.

• Preliminary economic assessment concluded that the HPLWR has the potential to be economically competitive. However, an accurate assessment can only be done after the HPLWR design has been fixed. A more accurate economic assessment may be performed after the conclusion of this project.

Sea-water desalination with nuclear and other energy sources: the EURODESAL project

This paper summarises our recent investigations undertaken as part of the EURODESAL project on nuclear desalination, currently being carried out by a consortium of four European, and one Canadian, industrials and two leading EU R&D organisations.Major achievements of the project, as discussed in this paper are:

• Coherent demonstration of the technical feasibility of nuclear desalination through the elaboration of coupling schemes for optimum cogeneration of electricity and water and by exploring the unique capabilities of the innovative nuclear reactors and desalination technologies.

• Verification that the integrated system design does not adversely affect nuclear reactor safety.

• Development of codes and methods for an objective economic assessment of the competitiveness and sustainability of proposed options through comparison, in European conditions, with fossil energy based systems.

Results obtained so far seem to be quite encouraging as regards the economical viability of nuclear desalination options.Thus, for example, specific desalination costs ($/m³ of desalted water) for nuclear systems, such as the AP-600 and the French PWR-900 (reference base case), coupled to multiple effect distillation (MED) or the reverse osmosis (RO) processes, are 30–60% lower than the desalination costs for fossil energy based systems, using pulverised coal and natural gas with combined cycle, at low discount rates and recommended fossil fuel prices. Even in the most unfavourable scenarios for nuclear energy (discount rate=10%, low fossil fuel costs) desalination costs with the nuclear reactors are 7–20% lower, depending upon the desalination capacities. Furthermore, with the advanced coupling schemes, utilising waste heat from nuclear reactors, the gains in specific desalination costs of nuclear systems are increased by another 2–15%, even without system and design optimisation. A preliminary evaluation shows that desalination costs with the GT-MHR, coupled to a MED process, could still be much lower than the above nuclear options for desalting capacities≤43 000 m³ per day. This is because its design intrinsically provides “virtually free” heat at ideal temperatures for desalination (80–100 °C).     

Modular HTR confinement/containment and the protection against aircraft crash

Does an HTR need a containment – pressure resistant – or is it possible – licensable – to have only a so-called confinement.The answer depends on both the results of the safety analysis of the accidents considered in the design and the acceptance by the licensing authorities and the public of a safety approach only based on severe core damage exclusion.The safety approach to be developed for modular HTRs must describe the application of the defence in depth principle for such reactors. Whatever the requirements on the last confinement barrier could be, a convincing demonstration of the exclusion of any severe core damage is needed, relying on exhaustive and bounding considerations of severe core damage initiators and the use of non-questionable arguments.The paper presents the containment issues for HTRs based on German experience background and considerations for modern modular HTR safety approach including beyond design situations.

• For the German HTRs (designed in the 80s), it could be shown in the licensing procedures in Germany that there was no need for a pressure retaining and gas tight containment to enclose radioactive nuclides released from the nuclear heat source. Instead, the confinement envelope acted in conjunction with other barriers to minimize the release of radioactive nuclides and the radiological impact on the environment.

• The confinement envelope consisted of the reactor building, a sub-atmospheric pressure system, a building pressure relief system, an HVAC systems isolation and a filtration system.

• During a major depressurization accident, unfiltered releases were discharged to the environment. The analyses results show that the environmental impact was far below the dose limits according to the German Radiological Protection Ordinance, even when the effect of filters was not taken into account.

• The demonstration strongly relied on the assumptions made for the source term definition, e.g. the fuel particles failure rates (under irradiation and during accidental conditions), the diffusion data, the dust data and the deposition/lift-off mechanisms.

• For modern modular HTRs, the last confinement barrier performances will have to be determined in accordance with the set of accidents to be considered in the design including internal and external hazards and the limits targeted for the public and the environment protection.

Further more the paper presents an analysis of effects of a deliberate crash of a large commercial airliner on a former German HTR design.     

The 2006 CHF look-up table

CHF look-up tables are used widely for the prediction of the critical heat flux (CHF). The CHF look-up table is basically a normalized data bank for a vertical 8 mm water-cooled tube. The 2006 CHF look-up table is based on a database containing more than 30,000 data points and provides CHF values at 24 pressures, 20 mass fluxes, and 23 qualities, covering the full range of conditions of practical interest. In addition, the 2006 CHF look-up table addresses several concerns with respect to previous CHF look-up tables raised in the literature. The major improvements of the 2006 CHF look-up table are:

• An enhanced quality of the database (improved screening procedures, removal of clearly identified outliers and duplicate data).

• An increased number of data in the database (an addition of 33 recent data sets).

• A significantly improved prediction of CHF in the subcooled region and the limiting quality region.

• An increased number of pressure and mass flux intervals (thus increasing the CHF entries by 20% compared to the 1995 CHF look-up table).

• An improved smoothness of the look-up table (the smoothness was quantified by a smoothness index).

A discussion of the impact of these changes on the prediction accuracy and table smoothness is presented. The 2006 CHF look-up table is characterized by a significant improvement in accuracy and smoothness.     

A look-up table for fully developed film-boiling heat transfer

An improved look-up table for film-boiling heat-transfer coefficients has been derived for steam–water flow inside vertical tubes. Compared to earlier versions of the look-up table, the following improvements were made:

• The database has been expanded significantly. The present database contains 77,234 film-boiling data points obtained from 36 sources.

• The upper limit of the thermodynamic quality range was increased from 1.2 to 2.0. The wider range was needed as non-equilibrium effects at low flows can extend well beyond the point where the thermodynamic quality equals unity.

• The surface heat flux has been replaced by the surface temperature as an independent parameter.

• The new look-up table is based only on fully developed film-boiling data.

• The table entries at flow conditions for which no data are available is based on the best of five different film-boiling prediction methods.

The new film-boiling look-up table predicts the database for fully developed film-boiling data with an overall rms error in heat-transfer coefficient of 10.56% and an average error of 1.71%. A comparison of the prediction accuracy of the look-up table with other leading film-boiling prediction methods shows that the look-up table results in a significant improvement in prediction accuracy.     

Methodology for the reliability evaluation of a passive system and its integration into a Probabilistic Safety Assessment

A methodology has been developed to evaluate the reliability of passive systems characterised by a moving fluid and whose operation is based on thermal–hydraulic (T-H) principles. The methodology includes:

• identification and quantification of the sources of uncertainties and determination of the important variables;

• propagation of the uncertainties through T-H models and assessment of T-H passive system unreliability;

• introduction of passive system unreliability in the accident sequence analysis.

Each step of the methodology is described and commented and a diagram of the methodology is presented. An example of passive system is presented with the aim to illustrate the possibilities of the methodology. This example is the Residual Passive heat Removal system on the Primary circuit (RP2), an innovating system supposed to be implemented on a 900 MWe Pressurized Water Reactor.     

European Validation of the Integral Code ASTEC (EVITA): First experience in validation and plant sequence calculations

The main objective of the European Validation of the Integral Code ASTEC (EVITA) project is to distribute the severe accident integral code ASTEC to European partners in order to apply the validation strategy issued from the VASA project (4th EC FWP). Partners evaluate the code capability through validation on reference experiments and plant applications accounting for severe accident management measures, and compare results with reference codes.The basis version V0 of ASTEC (Accident Source Term Evaluation Code)—commonly developed and basically validated by GRS and IRSN—was made available in late 2000 for the EVITA partners on their individual platforms. Users’ training was performed by IRSN and GRS. The code portability on different computers was checked to be correct. A “hot line” assistance was installed continuously available for EVITA code users. The actual version V1 has been released to the EVITA partners end of June 2002. It allows to simulate the front-end phase by two new modules:

• for reactor coolant system 2-phase simplified thermal hydraulics (5-equation approach) during both front-end and core degradation phases,

• for core degradation, based on structure and main models of ICARE2 (IRSN) reference mechanistic code for core degradation and on other simplified models.

Next priorities are clearly identified: code consolidation in order to increase the robustness, extension of all plant applications beyond the vessel lower head failure and coupling with fission product modules, and continuous improvements of users’ tools.As EVITA has very successfully made the first step into the intention to provide end-users (like utilities, vendors and licensing authorities) with a well validated European integral code for the simulation of severe accidents in NPPs, the EVITA partners strongly recommend to continue validation, benchmarking and application of ASTEC. This work will continue in Severe Accident Research Network (SARNET) in the 6th Framework Programme where ASTEC plays a key role as the reference European integral code.     

Objectives in developing the European pressurized water reactor (EPR)

In 1991, major German utilities and Electricité de France agreed to develop, together with Siemens and Framatome, the nuclear island for the next generation of nuclear power plants. This nuclear island design is based on German and French experience in the construction and operation of pressurized water reactors. The major step in the evolutionary European pressurized water reactor (EPR) design is the systematic inclusion of events beyond classic design events. The mitigation of core melt accidents by special means, such as reinforcement of the containment function, primary loop depressurization, hydrogen reduction and ultimate heat removal, is part of the design, in addition to a number of features which increase the reliability of plant operation and accident control. Nevertheless, utilities and designers are aware of the economic challenge facing the EPR by the need to compete with other nuclear power plant designs and fossil plants for electricity generation.     

The EPR overall approach for severe accident mitigation

The EPR design includes provisions to cope with severe accidents including core melt situations:

- Situations that would lead to large early releases such as containment bypass, high reactivity accidents, high-pressure core melt or global hydrogen detonation have to be prevented.

- All other situations, including low pressure core melt have to be mitigated in such a way that the corresponding radiological consequences would necessitate only very limited protective countermeasures in a relatively small area and for a limited time for the population living in the neighborhood of the power plant. This means that there would be no need for permanent relocation, no need for evacuation outside the immediate vicinity of the plant, limited sheltering and no long-term restrictions in food consumption.

To reach this objective, which is one of the Safety Authority's requirements, the EPR relies on a very robust containment and on various design measures intended to withstand extreme loads caused by a large range of internal events and external hazards. The deterministic method is used for the EPR safety demonstration, supplemented by probabilistic methods and appropriate R&D work.This paper outlines the major mitigating design features, summarizes the results of the Level 2 PSA and gives the main results of the evaluation of the radiological consequences of core melt on the environment.     

Characteristics of corium debris bed generated in large-scale fuel-coolant interaction experiments

The formation of corium debris as the result of fuel-coolant interaction (energetic or not) has been studied experimentally in the FARO and KROTOS facilities operated at JRC-Ispra between 1991 and 1999. Experiments were performed with 3–177 kg of UO₂–ZrO₂ and UO₂–ZrO₂–Zr melts, quenched in water at depth between 1 and 2 m, and pressure between 0.1 and 5.0 MPa. The effect of various parameters such as melt composition, system pressure, water depth and subcooling on the quenching processes, debris characteristics and thermal load on bottom head were investigated, thus, giving a large palette of data for realistic reactor situations.Available data related to debris coolability aspects in particular are:

• Geometrical configuration of the collected debris.

• Partition between loose and agglomerated (“cake”) debris.

• Particle size distribution with and without energetic interaction.

These data are synthesised in the present contribution.     

A multi-tank storage facility to effect power control in the PBMR power cycle

This article presents the concept of a storage facility used to effect power control in South Africa's PBMR power cycle. The concept features a multiple number of storage vessels whose purpose is to contain the working medium, helium, as it is withdrawn from the PBMR's closed loop power cycle, at low energy demand. This helium is appropriately replenished to the power cycle as the energy demand increases. Helium mass transfer between the power cycle and the storage facility, henceforth known as the inventory control system (ICS), is carried out by way of the pressure differential that exists between these two systems. In presenting the ICS concept, emphasis is placed on storage effectiveness; hence the discussion in this paper is centred on those features which accentuate storage effectiveness, namely:

• Storage vessel multiplicity;

• Unique initial pressures for each vessel arranged in a cascaded manner; and

• A heat sink placed in each vessel to provide thermal inertia.

Having presented the concept, the objective is to qualitatively justify the presence of each of the above-mentioned features using thermodynamics as a basis.     

Prediction of fuel channel—graphite gas-gap behaviour in RBMK reactors

To maintain thermal contact between the fuel assembly and the graphite moderator, RBMK design reactors employ graphite split rings, which are alternatively tight on the pressure tube or tight on the graphite brick central bore. The split in the graphite rings allows a helium/nitrogen gas mixture to flow up the fuel channel. This prevents oxidation of the graphite and can be sampled to detect pressure tube leaks. The initial clearance between the rings and pressure tube or graphite brick is approximately 2.7 mm (1.35 mm each side). Due to material property changes of the pressure tubes and graphite during operation of the reactor, the size of the clearance between the rings and the pressure tube/brick, called the “gas-gap”, varies. Closure of these gaps has been identified as a possible safety case issue by reactor designers and by independent reviews carried out as part of TACIS reviews and as part of the Ignalina Safety Analysis Report. The reasons for this are that gas-gap closure would cause the pressure tube to be tightly gripped by the graphite bricks via the split rings, which could lead to:

• Extra loading on the upper pressure tube zirconium/steel transition joint, particularly during shut down and emergency transients.

• Splitting of the graphite brick, leading to loss of thermal contact between the pressure tube and graphite. As approximately 5.6% of the heat in graphite-moderated reactor is generated within the moderator through neutron and gamma-heating, loss of thermal contact would result in higher graphite temperatures, accelerating the rate of graphite expansion and hence increasing the loading of the core radial restraint.

• Graphite debris may become lodged in inter-brick gaps, leading to increased axial pressure tube loading during shut down and emergency transients.

The authors have carried out deterministic assessments based on the Ignalina RBMK-1500 reactors in Lithuania, modelling the behaviour of the graphite under irradiation and have predicted graphite bore diameter changes that are in good agreement with the measurements of graphite bore diameters taken at Ignalina Nuclear Power Plant (NPP). A probabilistic model has been developed using the actual results of the deterministic calculations with non-linear graphite behaviour. Statistical analysis of the measurements of tube and graphite diameters taken from Units 1 and 2 at Ignalina NPP has been carried out. Further work has been carried out to try to determine the uncertainty inherent in the predictions of the gas-gap closure from the calculations. The overall objective of the studies is to aid prediction of the gas-gap closure process, and help to identify a suitable monitoring strategy for gas-gap closure that could be used for any RBMK reactor.     

Employing industrial standards in software engineering for W7X

The stellarator W7X is a large complex experiment designed for continuous operation and planned to be operated for about 20 years. Software support is highly demanded for experiment preparation, operation and data analysis which in turn induces serious non-functional requirements on the software quality like, e.g.:

• high availability, stability, maintainability vs.

• high flexibility concerning change of functionality, technology, personnel

• high versatility concerning the scale of system size and performance

These challenges are best met by exploiting industrial experience in quality management and assurance (QM/QA), e.g. focusing on top-down development methods, developing an integral functional system model, using UML as a diagramming standard, building vertical prototypes, support for distributed development, etc., which have been used for W7X, however on an ‘as necessary’ basis. Proceeding in this manner gave significant results for control, data acquisition, corresponding database-structures and user applications over many years.As soon as production systems started using the software in the labs or on a prototype the development activity demanded to be organized in a more rigorous process mainly to provide stable operation conditions. Thus a process improvement activity was started for stepwise introduction of quality assuring processes with tool support taking standards like CMMI, ISO-15504 (SPICE) as a guideline. Experiences obtained so far will be reported.We conclude software engineering and quality assurance has to be an integral part of systems engineering right from the beginning of projects and be organized according to industrial standards to be prepared for the challenges of nuclear fusion research.     

Feasibility study of the cut and weld operations by RH on the cooling pipes of ITER NB components

The maintenance operations of ITER NB components inside the vessel - Beam Line Components (BLC's) involve the removal of the faulty component, its transport to the hot cell as well as the reverse operations of transport of the repaired/new component and its reinstallation inside the vessel. Prior to the removal of the BLC's the cooling pipes must be detached from the component following a procedure that applies to the cutting of the pipes and subsequent welding when the component is re-installed. The purpose of this study, conducted in the framework of EFDA, is to demonstrate the feasibility of the cut and weld operations on the water pipes of the BLC's using fully remote handling techniques. Viable technologies for the cut and weld operations have been identified within the study; in particular the following aspects will be presented in the paper:

• Different strategies can be pursued in the detachment of the components depending on the number of cut and weld operations to be performed on the pipes. The selected strategy will impact on the procedure to be followed likewise on important aspects as the requirements of the flexible joints assembled on the pipes.

• The existing cutting techniques have been examined in the light of the remotely performed pipe cutting at the NB cell. Modifications of commercial tools have been proposed in order to adapt them to the BLC's pipes requirements. The debris produced during the cutting process must be controlled and collected, therefore a cleaning system has been integrated in the adapted cutting tool referred above.

• The existing welding techniques have been also examined and compared based on different criteria such as complexity, reliability, alignment tolerances, etc. TIG welding is the preferred technique as it stands out for its superior performance. The commercial tools identified need to be adapted to the NB environment.

• The alignment of the pipes is a critical issue concerning the remote welding. A proper alignment system has been proposed taking into account the pre-selected welding technique.