Reactor cooling systems thermal-hydraulic assessment of the ASTEC V1.3 code in support of the French IRSN PSA-2 on the 1300 MWe PWRs

The French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe PWRs. This PSA-2 study is relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, a wide-ranging series of comparisons with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe-accident scenarios. The present paper details 4 out of the 14 studied scenarios: a 12-in. cold leg Loss of Coolant Accident (LOCA), a 2-tube Steam Generator Tube Rupture (SGTR), a 12-in. Steam Line Break (SLB) and a total Loss of Feed Water scenario (LFW). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and compared to the CATAHRE 2 V2.5 results. The ASTEC results of the core degradation phase are also presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results.     

ASTEC application to in-vessel corium retention

This paper summarizes the work done in the SARNET European Network of Excellence on Severe Accidents (6th Framework Programme of the European Commission) on the capability of the ASTEC code to simulate in-vessel corium retention (IVR). This code, jointly developed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und Reaktorsicherheit mbH (GRS) for simulation of severe accidents, is now considered as the European reference simulation tool.First, the DIVA module of ASTEC code is briefly introduced. This module treats the core degradation and corium thermal behaviour, when relocated in the reactor lower head. Former ASTEC V1.2 version assumed a predefined stratified molten pool configuration with a metallic layer on the top of the volumetrically heated oxide pool. In order to reflect the results of the MASCA project, improved models that enable modelling of more general corium pool configurations were implemented by the CEA (France) into the DIVA module of the ASTEC V1.3 code.In parallel, the CEA was working on ASTEC modelling of the external reactor vessel cooling (ERVC). The capability of the ASTEC CESAR circuit thermal-hydraulics to simulate the ERVC was tested. The conclusions were that the CESAR module is capable of simulating this system although some numerical and physical instabilities can occur. Developments were then made on the coupling between both DIVA and CESAR modules in close collaboration with IRSN. In specific conditions, code oscillations remain and an analysis was made to reduce the numerical part of these oscillations. A comparison of CESAR results of the SULTAN experiments (CEA) showed an agreement on the pressure differences.The ASTEC V1.2 code version was applied to IVR simulation for VVER-440/V213 reactors assuming defined corium mass, composition and decay heat. The external cooling of reactor wall was simulated by applying imposed coolant temperature and heat transfer coefficient (HTC). The obtained results (pool temperatures, heat flux distribution, reactor wall ablation) were compared with available predictions of other codes. The agreement was correct, in particular on the shape and depth of ablation, as well as the maximum heat flux in case of a thick metallic layer, while ASTEC calculated a lower maximum heat flux for a thin metallic layer.     

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.

Progress on ruthenium release and transport under air ingress conditions

A particular concern in the event of a hypothetical severe accident is the potential release of highly radiotoxic fission product (FP) isotopes of ruthenium. The highest risk for a large quantity of these isotopes to reach the containment arises from air ingress following vessel melt-through. One work package (WP) of the source term topic of the EU 6th Framework Network of Excellence project SARNET is producing and synthesizing information on ruthenium release and transport with the aim of validating or improving the corresponding modelling in the European ASTEC severe accident analysis code. The WP includes reactor scenario studies that can be used to define conditions for new experiments.The experimental database currently being reviewed includes the following programmes:

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AECL experiments conducted on fission product release in air; results are relevant to CANDU loss of end-fitting accidents;

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VERCORS tests on FP release and transport conducted by CEA in collaboration with IRSN and EDF; additional tests may potentially be conducted in more oxidizing conditions in the VERDON facility;

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RUSET tests by AEKI investigating ruthenium transport with and without other FP simulants;

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Experiments by VTT on ruthenium transport and speciation in highly oxidizing conditions.

In addition to the above, at IRSN and at ENEA modelling of fission product release and of fuel oxidation is being pursued, the latter being an essential boundary condition influencing ruthenium release.Reactor scenario studies have been carried out at INR, EDF and IRSN: calculations of air ingress scenarios with respectively ICARE/CATHARE V2; SATURNE-MAAP; and ASTEC codes provided first insights of thermal-hydraulic conditions that the fuel may experience after lower head vessel failure.This paper summarizes the status of this work and plans for the future.     

Sustainable integration of EU research in severe accident phenomenology and management

In order to optimise the use of the available means and to constitute sustainable research groups in the European Union, the Severe Accident Research NETwork of Excellence (SARNET) has gathered, between 2004 and 2008, 51 organizations representing most of the actors involved in severe accident (SA) research in Europe plus Canada. This project was co-funded by the European Commission (EC) under the 6th Euratom Framework Programme. Its objective was to resolve the most important pending issues for enhancing, in regard of SA, the safety of existing and future nuclear power plants (NPPs).SARNET tackled the fragmentation that existed between the national R&D programmes, in defining common research programmes and developing common computer codes and methodologies for safety assessment. The Joint Programme of Activities consisted in:

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Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents;

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Harmonizing and re-orienting the research programmes, and defining new ones;

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Analyzing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena;

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Developing the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA) by capitalizing in terms of physical models the knowledge produced within SARNET;

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Developing scientific databases, in which the results of research experimental programmes are stored in a common format;

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Developing a common methodology for probabilistic safety assessment of NPPs;

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Developing short courses and writing a text book on severe accidents for students and researchers;

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Promoting personnel mobility amongst various European organizations.

This paper presents the major achievements after four and a half years of operation of the network, in terms of knowledge gained, of improvements of the ASTEC reference code, of dissemination of results and of integration of the research programmes conducted by the various partners.Most initial objectives were reached but the continuation of the SARNET network, co-funded by EC in the 7th Framework Programme (SARNET2 project that started in April 2009 for 4 years), will consolidate the first assets and focus mainly on the highest priority pending issues as determined during the first period. The objective will be also to make the network evolve towards a complete self-sustainability.     

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.

The EPR: A clear step forward in dose reduction and radiation protection

For the European Pressurized Water Reactor (EPR) a large effort was made to improve the plant design with respect to radiation protection using the experience gained during the design of former generations of pressurized water reactor (PWR) in France and Germany, and their current operation. Keeping the radiation exposure of personnel to an acceptable level is one of the main objectives of the EPR design. Both the individual and the collective doses are considered.Internationally comparable limits based on recommendations of the International Commission on Radiological Protection (ICRP) have been established for individual doses. These limits describe the framework within which the individual dose shall be kept as low as possible, applying the principles:

• Justification: No practice involving exposures to radiation should be adopted unless it produces sufficient benefit to the exposed individuals or to society to offset the radiation detriment it causes.

• Optimization: In relation to any particular source within a practice, the magnitude of individual doses, the number of people exposed, and the likelihood of incurring exposures where these are not certain to be received should all be kept as low as reasonably achievable (ALARA), economic and social factors being taken into account.

• Limitation: The exposure of individuals resulting from the combination of all the relevant practices is subject to dose limits.

The paper describes the design provision and measures introduced in the plant design to achieve the above described goals.They are in essence:

• measures to avoid or to reduce sources of radiation;

• layout aspects;

• provisions made in the component design with respect to ease of operation and maintenance management;

• improved possibilities of decontamination;

• use of operating experience for design improvements.

The radiation protection layout principles compiled on the basis of safe operating experience gained from the existing pressurized water reactors in France and Germany are used to develop an improved plant design with respect to radiation protection aspects and dose optimization.Summary: The European Pressurized Water Reactor is an evolutionary third-generation pressurized water reactor with a rating in the 1600 MWe class. Its development was started in 1992 by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, now AREVA NP. Being the product of intense bilateral cooperation the EPR combines the technological accomplishments of the world's two leading PWR product lines—the French N4 reactors in operation at Chooz and Civeaux and the Konvoi reactors in operation at Neckarwestheim, Emsland and Isar in Germany. From the very start, development of the EPR was focused on improving plant safety and economics even further and also a large effort was made to improve the plant design with respect to radiation protection. Keeping the doses received by operating and maintenance personnel to a level far below the limiting values was one of the main objectives of the EPR design. Both the individual and the collective doses are considered in this article.     

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.     

Air oxidation of Zircaloy-4 in the 600-1000 °C temperature range: Modeling for ASTEC code application

Progress in the treatment of air oxidation of zirconium in severe accident (SA) codes are required for a reliable analysis of severe accidents involving air ingress. Air oxidation of zirconium can actually lead to accelerated core degradation and increased fission product release, especially for the highly-radiotoxic ruthenium. This paper presents a model to simulate air oxidation kinetics of Zircaloy-4 in the 600-1000 °C temperature range. It is based on available experimental data, including separate-effect experiments performed at IRSN and at Forschungszentrum Karlsruhe. The kinetic transition, named “breakaway”, from a diffusion-controlled regime to an accelerated oxidation is taken into account in the modeling via a critical mass gain parameter. The progressive propagation of the locally initiated breakaway is modeled by a linear increase in oxidation rate with time. Finally, when breakaway propagation is completed, the oxidation rate stabilizes and the kinetics is modeled by a linear law. This new modeling is integrated in the severe accident code ASTEC, jointly developed by IRSN and GRS. Model predictions and experimental data from thermogravimetric results show good agreement for different air flow rates and for slow temperature transient conditions.     

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.

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.     

Effect of steam environment on severe core damage behaviour for VVER-1000 with the ASTEC V1 code

Severe accident studies for very low frequency events for VVER-1000 (V320) are carried out to estimate in-vessel damage progression under steam-rich and starved conditions. The analyses with code ASTEC, jointly developed by IRSN (France) and GRS, Germany), have shown the influence of steam environment on core heat-up followed by material relocation, hydrogen production, vessel failure and aerosol generation along with release to containment. Hydro-accumulator injection for studied transients also gives rise to a steam-rich environment enhancing the material oxidation depending on the injection time and period. The generated information along with PSA-Level 2 is helpful to decide Plant Damage State (PDS) and fruitfully develop accident management strategies for the plant.     

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.     

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.     

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.     

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.     

Thermal-hydraulics analyses of ELSY lead fast reactor with open square core option

Among the next generation of nuclear reactors, well-known as Gen-IV, the LFR (Lead Fast Reactor) is one of the most promising advanced reactor able to comply the principles of sustainability, economics, safety and proliferation resistance. The ELSY project (European Lead-cooled SYstem), funded by the 6th European Framework Programme, aims at investigating the technical/economical feasibility of a high power lead fast reactor with waste transmutation capability.Several innovative design solutions have been proposed at system level and some of them regard the core region with open-assemblies that represents the reference option. To support the design phase and the safety assessment of ELSY, the thermal-hydraulic code RELAP5, modified to treat heavy liquid metals, has been taken into account.The paper deals with the development of the ELSY thermal-hydraulic and point kinetic model for RELAP5 focusing the attention at core assembly level to verify that the temperatures at nominal power conditions stay within the safety limits both in Beginning-of-Cycle (BOC) and in End-of-Cycle (EOC) conditions. Moreover, in order to have a first evaluation of the system behavior in accidental conditions, an Unprotected Loss-of-Flow Accident (ULOF) simulation at BOC has been analysed and discussed.

List of acronyms

ADS

Accelerator Driven System

BOC

Beginning-of-Cycle

DEC

Design Extension Condition

EFIT

European Facility for Industrial Transmutation

ELSY

European Lead-cooled SYstem

EOC

End-of-Cycle

FA

Fuel Assembly

FP6

6th Framework Programme

Gen-IV

Generation IV (four)

GESA

Gepulste ElektronenStrahlanlage (Pulsed Electron Beam Facility)

HX

Heat eXchanger

IC

Isolation Condenser

LFR

Lead Fast Reactor

LMFR

Liquid Metal Fast Reactor

PDS-XADS

Preliminary Design Studies on eXperimental ADS

RVACS

Reactor Vessel Air Cooling System

ULOF

Unprotected Loss-of-Flow

W-DHR

Water Decay Heat Removal